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1

Brookhaven Graphite Research Reactor | Environmental Restoration...  

NLE Websites -- All DOE Office Websites (Extended Search)

Brookhaven Graphite Research Reactor(BGRR) BGRR Overview BGRR Complex Description Decommissioning Decision BGRR Complex Cleanup Actions BGRR Documents BGRR Science &...

2

Cleanup at the Brookhaven Graphite Research Reactor  

NLE Websites -- All DOE Office Websites (Extended Search)

Graphite Research Reactor placeholder Remotely operated robot known as a BROKK manipulator In April 2005, the Department of Energy (DOE), the Lab, and the regulatory agencies...

3

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

NLE Websites -- All DOE Office Websites (Extended Search)

Brookhaven Graphite Research Reactor Documents Brookhaven Graphite Research Reactor Documents Feasibility Study (PDF) Proposed Remedial Action Plan (PDF) Record of Decision (PDF) RD/RA Work Plan for the BGRR Pile (PDF) RD/RA Work Plan for the Bioshield (PDF) RD/RA Work Plan for the BGRR Cap (PDF) Brookhaven Graphite Research Reactor Explanation of Significant Differences (PDF) (4/12) NYSDEC Approval Letter for BGRR ESD (PDF) (5/12) USEPA Approval Letter for BGRR ESD (PDF) (6/12) DOE BGRR ESD Transmittal Letter (PDF) (7/12) Remedial Design Implementation Report (PDF) (12/11) Completion Reports Removal of the Above-Ground Ducts and Preparation of the Instrument House (708) for Removal (PDF) - April 2002 Below-Ground Duct Outlet Air Coolers, Filters and Primary Liner Removal (PDF) - April 2005 Canal and Deep Soil Pockets Excavation and Removal (PDF) - August

4

Brookhaven Lab Completes Decommissioning of Graphite Research Reactor:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Brookhaven Lab Completes Decommissioning of Graphite Research Brookhaven Lab Completes Decommissioning of Graphite Research Reactor: Reactor core and associated structures successfully removed; waste shipped offsite for disposal Brookhaven Lab Completes Decommissioning of Graphite Research Reactor: Reactor core and associated structures successfully removed; waste shipped offsite for disposal September 1, 2012 - 12:00pm Addthis The Brookhaven Graphite Research Reactor’s bioshield, which contains the 700-ton reactor core, is shown prior to decommissioning. The Brookhaven Graphite Research Reactor's bioshield, which contains the 700-ton reactor core, is shown prior to decommissioning. Pictured here is the Brookhaven Graphite Research Reactor, where major decommissioning milestones were recently reached after the remaining radioactive materials from the facility’s bioshield were shipped to a licensed offsite disposal facility.

5

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

NLE Websites -- All DOE Office Websites (Extended Search)

Why Was the BGRR Decommissioned? Why Was the BGRR Decommissioned? BGRR The Brookhaven Graphite Research Reactor (BGRR) at Brookhaven National Laboratory (BNL) was decommissioned to ensure the complex is in a safe and stable condition and to reduce sources of groundwater contamination. The BGRR contained over 8,000 Curies of radioactive contaminants from past operations consisting of primarily nuclear activation products such as hydrogen-3 (tritium) and carbon-14 and fission products cesium-137 and strontium-90. The nature and extent of contamination varied by location depending on historic uses of the systems and components and releases, however, the majority of the contamination (over 99 percent) was bound within the graphite pile and biological shield. Radioactive contamination was identified in the fuel handling system deep

6

Brookhaven Graphite Research Reactor Workshop | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Services » Site & Facility Restoration » Deactivation & Services » Site & Facility Restoration » Deactivation & Decommissioning (D&D) » D&D Workshops » Brookhaven Graphite Research Reactor Workshop Brookhaven Graphite Research Reactor Workshop The Brookhaven Graphite Research Reactor (BGRR) was the first reactor built in the U.S. for peacetime atomic research following World War II. Construction began in 1947 and the reactor started operating in August 1950. In the next 18 years, an estimated 25,000 scientific experiments were carried out at the BGRR using neutrons produced in the facility's 700-ton graphite core, made up of more than 60,000 individual graphite blocks. The BGRR was placed on standby in 1968 and then permanently shut down as the next-generation reactor, the High Flux Beam Reactor (HFBR), was

7

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

NLE Websites -- All DOE Office Websites (Extended Search)

- Cleanup Actions - Cleanup Actions Since the Brookhaven Graphite Research Reactor (BGRR) was shut down in 1968, many actions have been taken as part of the complex decommissioning. The actions undertaken throughout the BGRR complex ensure that the structures that remain are in a safe and stable condition and prepared it for long-term surveillance and maintenance. Regulatory Requirements The decommissioning of the BGRR was conducted under the federal Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). In 1992, an Interagency Agreement (PDF) among the DOE, the U.S. Environmental Protection Agency (EPA) and the New York State Department of Environmental Conservation (NYSDEC) became effective. The IAG provided the overall framework for conducting environmental restoration activities at

8

Brookhaven Graphite Research Reactor | Environmental Restoration...  

NLE Websites -- All DOE Office Websites (Extended Search)

of multiple structures and systems that were necessary to operate and maintain the reactor. The most recognizable features of the complex include the Building 701 officeshigh...

9

In accounts of seminal neutron research at ORNL's Graphite Reactor,  

E-Print Network (OSTI)

requested permission to set up an X-ray diffractometer he had brought from the University of Chicago for his of the Graphite Reactor. "I was a student at the University of Chicago in 1942 when Enrico Fermi was doing his. 25 Scrooge (OR Playhouse) Nov. 27 Football: UT vs. Kentucky Dec. 4 Fiddler on the Roof Dec. 11 Best

10

STORED ENERGY: GROWTH AND ANNEALING STATUS OF GRAPHITE MODERATOR IN THE BNL RESEARCH REACTOR. Final Report  

SciTech Connect

The present sthtus, past annealing procedures and experiences, future annealing procedures, annealing sehedule, revised annealing procedure (1958), procedure for combating a graphite fire in fuel channel, high-temperature stored energy, and graphite burning experiments are reportcd for the BNL Research Reactor. The following subjccts are discussed in the appendixes: control of radiation damage in a graphitc reactor; annealing of graphite moderator structure in the BNL; annealing operation in BNL graphite reactor; effect of pile radiation on mechanical and other properties of graphite; neutron sensing instrumentation; instrumentation for sensing fuel failures; thermocouple pattern for enriched fuel loading; environmental hazard from a molten fuel element; retention of volatile flssion products on filters; retention of volatile fission products on water tube coolers; retention of volatile fission products in molten fuel plates; and release of the lowtemperature stored energy in the BEPO Pile. (W.L.H.)

1959-10-31T23:59:59.000Z

11

Innovative Graphite Removal Technology for Graphite Moderated Reactor Decommissioning  

Science Conference Proceedings (OSTI)

This report defines a trial program to support the development of a new concept for the removal of reactor graphite by remote in-situ size reduction and vacuum transfer, known as nibble-and-vacuum. This new approach to graphite retrieval has significant potential for simplifying the decommissioning process of graphite moderated reactors. It produces graphite gravel, which has potential as feedstock for processes such as gasification/steam reforming. This report includes definition of the trial program, t...

2010-09-28T23:59:59.000Z

12

Graphite Reactor | ornl.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Graphite Reactor Graphite Reactor 'In the early, desperate days of World War II, the United States launched the top-secret, top-priority Manhattan Project...' In the early, desperate days of U.S. involvement in World War II, American scientists began to fear that the German discovery of uranium fission in 1939 might enable the Nazis to develop a super bomb. Afraid of losing this crucial race, the United States launched the top-secret, top-priority Manhattan Project. The plan was to create two atomic weapons-one fueled by plutonium, the other by enriched uranium. Hanford, Washington, was selected as the site for plutonium production, but before large reactors could be built there, a pilot plant was necessary to prove the feasibility of scaling up from laboratory experiments. A secluded, rural area near Clinton, Tennessee, was

13

X-10 Graphite Reactor | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor When President Roosevelt in December 1942 authorized the Manhattan Project, the Oak Ridge site in eastern Tennessee had already been obtained and plans laid for an air-cooled experimental pile, a pilot chemical separation plant, and support facilities. The X-10 Graphite Reactor, designed and built in ten months, went into operation on November 4, 1943. The X-10 used neutrons emitted in the fission of uranium-235 to convert uranium-238 into a new element, plutonium-239. The reactor consists of a huge block of graphite, measuring 24 feet on each side, surrounded by several feet of high-density concrete as a radiation shield. The block is pierced by 1,248 horizontal diamond-shaped channels in

14

ADMINISTRATION OF ORNL RESEARCH REACTORS  

SciTech Connect

Organization of the ORNL Operations division for administration of the Oak Ridge Research Reactor, the Low Intensity Testing Reactor, and the Oak Ridge Graphite Reactor is described. (J.R.D.)

Casto, W.R.

1962-08-20T23:59:59.000Z

15

US graphite reactor D&D experience  

Science Conference Proceedings (OSTI)

This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE).

Garrett, S.M.K.; Williams, N.C.

1997-02-01T23:59:59.000Z

16

History of Research Reactors at Brookhaven  

NLE Websites -- All DOE Office Websites (Extended Search)

History of Research Reactors at Brookhaven History of Research Reactors at Brookhaven Brookhaven National Laboratory has three nuclear reactors on its site that were used for scientific research. The reactors are all shut down, and the Laboratory is addressing environmental issues associated with their operations. photo of BGRR Brookhaven Graphite Research Reactor - Beginning operations in 1950, the graphite reactor was used for research in medicine, biology, chemistry, physics and nuclear engineering. One of the most significant achievements at this facility was the development of technetium-99m, a radiopharmaceutical widely used to image almost any organ in the body. The graphite reactor was shut down in 1969. Parts of it have been decommissioned, with the remainder to be addressed by 2011. More history

17

FUEL PROGRAMMING FOR SODIUM GRAPHITE REACTORS  

SciTech Connect

The effect of fuel programming, i.e., the scheme used for changing fuel in a core, on the reactivity and specific power of a sodium graphite reactor is discussed Fuel programs considered Include replacing fuel a core-load at a time or a radial zone at a time, replacing fuel to manutain the same average exposure of fuel elements throughout the core, and replacing and transferring fuel elements to maintain more highly exposed fuel in the center or at the periphery of the core. Flux and criticality calculations show the degree of power flattening and the concurrent decrease in effective multiplication which results from maintaining more exposed fuel toward the core center. Corverse effects are shown for the case of maintaining more exposed fuel near the core periphery. The excess reactivity which must be controlled in the various programs is considered. Illustrative schedules for implementing each of these programs in an SGR are presented. (auth)

Connolly, T.J.

1959-10-15T23:59:59.000Z

18

A safety assessment of the use of graphite in nuclear reactors licensed by the US NRC  

DOE Green Energy (OSTI)

This report reviews existing literature and knowledge on graphite burning and on stored energy accumulation and releases in order to assess what role, if any, a stored energy release can have in initiating or contributing to hypothetical graphite burning scenarios in research reactors. It also addresses the question of graphite ignition and self-sustained combustion in the event of a loss-of-coolant accident (LOCA). The conditions necessary to initiate and maintain graphite burning are summarized and discussed. From analyses of existing information it is concluded that only stored energy accumulations and releases below the burning temperature (650/sup 0/C) are pertinent. After reviewing the existing knowledge on stored energy it is possible to show that stored energy releases do not occur spontaneously, and that the maximum stored energy that can be released from any reactor containing graphite is a very small fraction of the energy produced during the first few minutes of a burning incident. The conclusions from these analyses are that the potential to initiate or maintain a graphite burning incident is essentially independent of the stored energy in the graphite, and depends on other factors that are unique for these reactors, research reactors, and for Fort St. Vrain. In order to have self-sustained rapid graphite oxidation in any of these reactors, certain necessary conditions of geometry, temperature, oxygen supply, reaction product removal, and a favorable heat balance must be maintained. There is no new evidence associated with either the Windscale Accident or the Chernobyl Accident that indicates a credible potential for a graphite burning accident in any of the reactors considered in this review.

Schweitzer, D.G.; Gurinsky, D.H.; Kaplan, E.; Sastre, C.

1987-09-01T23:59:59.000Z

19

Brookhaven Medical Research Reactor  

NLE Websites -- All DOE Office Websites (Extended Search)

Medical Research Reactor BMRR The last of the Lab's reactors, the Brookhaven Medical Research Reactor (BMRR), was shut down in December 2000. The BMRR was a three megawatt...

20

BNL | Our History: Reactors as Research Tools  

NLE Websites -- All DOE Office Websites (Extended Search)

> See also: Accelerators > See also: Accelerators Brookhaven History: Using Reactors as Research Tools BGRR Brookhaven Graphite Research Reactor The Brookhaven Graphite Research Reactor (BGRR) was the Laboratory's first big machine and the first peace-time reactor built in the United States following World War II. The reactor's primary mission was to produce neutrons for scientific experimentation and to refine reactor technology. At the time, the BGRR could accommodate more simultaneous experiments than any other reactor. Scientists and engineers from every corner of the U.S. came to use the reactor, which was not only a source of neutrons for experiments, but also an excellent training facility. Researchers used the BGRR's neutrons as tools for studying atomic nuclei and the structure of solids, and to investigate many physical, chemical and

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

Research reactors - an overview  

SciTech Connect

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

West, C.D.

1997-03-01T23:59:59.000Z

22

Experimental Determination of the Effect of Reactor Radiation on the Thermal Conductivity of Uranium-Impregnated Graphite  

SciTech Connect

Experiments are described in which the change in thermal conductivity of U-impregnated graphite under neutron irradiation was measured. Thermal resistivities relative to the thermal resistivity of undamaged impregnated graphite are reorted as functions of exposure. From applications of the expermental results to the North American Aviation low-power research reactor the peak tem. of the core is determined for a given reactor power and time of operation.

Hetrick, D.L.; McCarty, W.K.; Steele, G.N.; Brown, M.S.; Clark, E.V.; Holmes, F.R.; Howard, D.F.; McElroy, W.N.; Shields, B.L.

1953-01-06T23:59:59.000Z

23

Basic and Applied Science Research Reactors - Reactors designed...  

NLE Websites -- All DOE Office Websites (Extended Search)

BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th...

24

Heavy Water and Graphite Reactors - Reactors designed/built by...  

NLE Websites -- All DOE Office Websites (Extended Search)

experiments, necessary to achieve higher precision for the determination of reactor power distribution patterns, effect of non-uniform void distributions, kinetic behavior,...

25

THE SODIUM GRAPHITE REACTOR: TOMMORROW'S POWER PLANT  

SciTech Connect

A description is given of the Advanced Sodium Graphite Reactor Power Plant, including the reactor, heat transfer systems, generatirg plant, control systems, and the economics of producing 256 Mw(e). The safety of this design is due to its unusually low operating pressure, absence of chemically incompatible materials in the core, and excellent stability under atatic and dynamic conditions. The reactor is being constructed at Hallam, Nebraska, at a probable cost of 0 to 0/kw, exclusive of the first core costs. The 151 fuel elements of uranium carbide are enriched to 2.75 at.% U/sup 235/ and clad in stainless steel. The average thermal neutron flux in the fuel is 8 x 10/sup 13/ n/cm/sup 2/sec. (B.O.G.)

Beeley, R.J.; Lowell, E.G.; Polak, H.; Renard, J.

1960-04-25T23:59:59.000Z

26

Advanced Nuclear Research Reactor  

SciTech Connect

This report describes technical modifications implemented by INVAP to improve the safety of the Research Reactors the company designs and builds.

Lolich, J.V.

2004-10-06T23:59:59.000Z

27

Design guide for category VI reactors: air-cooled graphite reactors  

SciTech Connect

The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned air-cooled graphite reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC).

Brynda, W.J.; Karol, R.; Powell, R.W.

1979-02-01T23:59:59.000Z

28

CORE PARAMETER STUDY FOR A 300-MW SODIUM GRAPHITE REACTOR  

SciTech Connect

A core parameter study of the operating costs was performed for a 300- Mwe sodium graphite reactor, a scale-up of the Hallam Power Reactor. The results of the study indicate that the core design is nsar optimum and that core modifications would reduce the power costs by less than 5%. The lattice spacing, fuel rod diameter, and sodium flow can be varied within a rather broad range without significant changes in power generation costs. The effect of the fuel cladning thickness is more significant; fuel cycle costs can be reduced if stainless steel canning is replaced with zirconium canning. Use of UC in place of uraniummolybdenum fuel would also permit cost reductions. (D.L.C.)

Corcoran, W.P.

1959-10-22T23:59:59.000Z

29

Foreign Research Reactor/Domestic Research Reactor Receipt Coordinator...  

National Nuclear Security Administration (NNSA)

Research ReactorDomestic Research Reactor Receipt Coordinator, Savannah River Nuclear Solutions | National Nuclear Security Administration Our Mission Managing the Stockpile...

30

Graphite Leaching  

Science Conference Proceedings (OSTI)

The graphite moderators of retired gas-cooled nuclear reactors present a difficult challenge during demolition activities. As part of the EPRI graphite initiative on the technical issues involved in the management and disposal of irradiated nuclear graphite, this report examines the international data on leaching of radioactive isotopes from graphite, relevant to the decommissioning of graphite-moderated reactors.

2008-05-21T23:59:59.000Z

31

THE SODIUM GRAPHITE REACTOR POWER PLANT FOR CPPD  

SciTech Connect

The plant arrangement, component design, and the functions of various systems are described and illustrated. Relative estimated costs of the systems and major components are indicated. The reactor core is designed around requiremouts for 254 thermal megawatts, 950 deg F maximum sodium temperature, stainless steel clad graphite moderator blocks, and low enrichment (0.015 to 0.04 U/sup 235/) uranium fuel elements. The fuel cycle is described for the possible fuel elements. The fuel cost factors are discussed. Burn-up limitations encountered for metallic fuel in the SGR temperature range indicate UO/sub 2/ the more desirable choice. The estimated cost of electrical energy associated with the UO/sub 2/ fuel is given. (auth)

Olson, R.L.; Gerber, R.C.; Gordon, R.B.; Ross-Clunis, H.A.; Stolz, J.F.

1958-10-31T23:59:59.000Z

32

Prerequisite requirements for higher graphite temperature limits and/or nitrogen atmosphere at all reactors  

SciTech Connect

The graphite temperature limits specified in the process standards-are being closely approached or have limited power levels at B, D, DR, and reactors. An increase of approximately 50--100 C in graphite temperature limits at these reactors would permit year-around operation on bulk outlet temperature limits. There has been considerable recent interest in extending to all reactors the use of nitrogen as a reactor gas constituent. With present graphite temperature limits, significant benefit from nitrogen usage will not be obtained during the winter months at reactors which are graphite temperature limited with approximately 100 per cent helium. However, during the summer months when bulk outlet temperature limitations result in graphite temperatures considerably below the maximum permissible graphite temperatures, the substitution of nitrogen for carbon dioxide could be of value. Under these circumstances, both a reduction in helium usage and a reduction in enrichment costs could result. With an increase in permissible graphite temperature limits, year-around benefit from reduced helium usage and reduced enrichment costs would be possible. To meet the Pu-240 specification with higher graphite temperatures however, would require a reduction in current goal discharge exposures with resulting increased fuel and burnout costs. Additionally, the incentives for higher graphite temperatures are very sensitive to both the mode of operation and to the assumed product worth. An economic study by the Process Analysis Unit discussing in detail the factors influencing the incentive for higher graphite temperature will be published in the near future. At a meeting among the staff several weeks ago, the prerequisite requirements for permitting higher graphite temperatures and/or nitrogen usage at all reactors were discussed. It is the purpose of this letter to briefly outline the prerequisite steps considered necessary to achieve these goals.

Graves, S.M.

1962-02-26T23:59:59.000Z

33

AN ADVANCED SODIUM-GRAPHITE REACTOR NUCLEAR POWER PLANT  

SciTech Connect

An advanced sodium-cooled, graphite-moderated nuclear power plant is described which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency. Steam is generated at 2400 psig, superheated to 1050 deg F and, after partial expansion in the turbine, reheated to 1000 deg F. Net thermal efficiency of the plant is 42.3%. In a plant sized to produce a net electrical output of 256 Mw, the estimated cost is 8232/kw. Estimated cost of power generation is 6.7 mills/kwh. In a similar plant with a net electrical output of 530 Mw, the estimated power generating cost is 5.4 mills/ kwh. Most of the components of the plant are within the capability of current technology. The major exception is the fuel material, uranium carbide. Preliminary results of the development work now in progress indicate that uranium carbide would be an excellent fuel for high-temperature reactors, but temperature and burnup limitation have yet to be firmly established. Additional development work is also required on the steam generators. These are the single-barrier type similar to those which will be used in the Enrico Fernri Fast Breeder Reactor plant but produce steam at higher pressure and temperature. Questions also remain regarding the use of nitrogen as a cover gas over sodium at 1200 deg F and compatibility of the materials used in the primary neutron shield. All of these questions are currently under investigation. (auth)

Churchill, J.R.; Renard, J.

1960-03-15T23:59:59.000Z

34

Chicago Pile reactors create enduring research legacy - Argonne's  

NLE Websites -- All DOE Office Websites (Extended Search)

Chicago Pile reactors create enduring research Chicago Pile reactors create enduring research legacy About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

35

Gaps at the periphery of N Reactor graphite stack: A photographic survey  

Science Conference Proceedings (OSTI)

Miniature camera techniques were used to examine regions at the perimeter of the N Reactor graphite stack. Extensive coverage in previously unexplored regions was obtained. Gaps at the top of the unit show no visible evidence of graphite growth related distortion or displacement. Debris apparently associated with leaks in the subcritical monitors is observed but there are no indications that the material adversely affects the functions of safety, cooling, or monitoring systems. Gaps inside the right primary shield reveal graphite cooling tubes and other components in undisturbed condition. The only changes since construction are a thin layer of graphite dust from earlier rod channel renovation and oxide films on steel components. In the steam vent between moderator and reflector graphite, displaced and broken graphite blocks are observed. Again, no functional impairment of rod channels or other systems is observed. There is no evidence of distortion or deterioration of the graphite cooling tubes. The surveillance summarized here explores many previously hidden recesses. Some findings were unpredicted but none have given cause to change the safety status of the ball channel or control rod systems or to initiate any new safety enhancement measures. This report establishes a baseline for visual surveillance in perimeter regions of the N Reactor stack if operations resume. 10 refs., 22 figs., 1 tab.

Woodruff, E.M.; Weber, J.R.; Sinclair, R.B.

1988-09-01T23:59:59.000Z

36

EVALUATION OF CALANDRIA, THIMBLE, AND CANNED-MODERATOR CONCEPTS FOR SODIUM GRAPHITE REACTORS  

SciTech Connect

In efforts to improve the neutron economy and lower the capital costs of sodium graphite reactors, several methods of separating the sodium and graphite were investigated including the calandria, the thimble, and the canned moderator reactors. An analysis including nuclear, heat transfer, and economic comparisons was made of these SGR concepts. Based upon neutron economy and feasibility of core fabrication, the calandria concept appears to offer the greatest potential for improvement in 8GR design. The thimble concept provides some improvement in neutron economy but introduced numerous problems requiring developmental work. (auth)

Reed, G.L.

1960-06-10T23:59:59.000Z

37

100-MW NUCLEAR POWER PLANT UTILIZING A SODIUM COOLED, GRAPHITE MODERATED REACTOR  

SciTech Connect

The conceptual design of a 100 Mw(e) nuclear power plant is described. The plant utilized a sodium-cooled graphite-moderated reactor with stainless- steel clad. slightiy enriched UO/sub 2/ fuel. The reactor is provided with three main coolant circuits, and the steam cycle has three stages of regenerative heating. The plant control system allows automatic operation over the range of 20 to 100% load, or manual operation at all loads. The site, reactor, sodium systems, reactor auxiliaries, fuel handling, instrumentation, turbine-generator, buildings. and safety measures are described. Engineering drawings are included. (W.D.M.)

1958-02-28T23:59:59.000Z

38

Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics  

E-Print Network (OSTI)

Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermal neutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermal neutron diffusion length in graphite is evaluated at $50.85 \\pm 0.3cm$ using COMSOL Multiphysics and $50.95 \\pm 0.5cm$ using MCNP. Finally ...

Mirfayzi, S R

2013-01-01T23:59:59.000Z

39

Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities  

E-Print Network (OSTI)

Small graphite-moderated and gas-cooled reactors have been around since the beginning of the atomic age. Though their existence in the past has been associated with nuclear weapons programs, they are capable of being used in civilian power programs. The simpler design constraints associated with this type of reactor would make them ideal for developing nations to bolster their electricity generation and help promote a greater standard of living in those nations. However, the same benefits that make this type of reactor desirable also make it suspicious to the international community as a possible means to shorten that state?s nuclear latency. If a safeguards approach could be developed for a fuel cycle featuring one of these reactors, it would ease the tension surrounding their existence and possibly lead to an increased latency through engineered barriers. The development of this safeguards approach follows a six step procedure. First, the fuel cycle was analyzed for the types of facilities found in it and how nuclear material flows between facilities. The goals of the safeguards system were established next, using the normal IAEA standards for the non-detection and false alarm probabilities. The 5 MWe Reactor was modeled for both plutonium production and maximum power capacity. Each facility was analyzed for material throughput and the processes that occur in each facility were researched. Through those processes, diversion pathways were developed to test the proposed safeguards system. Finally, each facility was divided into material balance areas and a traditional nuclear material accountancy system was set up to meet the established safeguards goals for the facility. The DPRK weapons program is a great example of the type of fuel cycle that is the problem. The three major facilities in the fuel cycle, the Fuel Fabrication Facility, the 5 MWe Reactor, and the Radiochemical Laboratory, can achieve the two goals of safeguards using traditional methods. Each facility can be adequately safeguarded using methods and practices that are relatively inexpensive and can obtain material balance periods close to the timeliness limits set forth by the IAEA. The Fuel Fabrication Facility can be safeguarded at both its current needed capacity and its full design capacity using inexpensive measurements. The material balance period needed for both capacities are reasonable. For the 5 MWe reactor, plutonium production is simulated to be 6.7 kg per year and is on the high side of estimates. The Radiochemical Laboratory can also be safeguarded at its current capacity. In fact, the timeliness goal for the facility dictates what the material balance period must be for the chosen set of detectors which make it very reasonable.

Rauch, Eric B.

2009-05-01T23:59:59.000Z

40

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors  

Science Conference Proceedings (OSTI)

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

Rosenthal, Murray Wilford [ORNL

2009-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
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41

Research on Very High Temperature Gas Reactors  

Science Conference Proceedings (OSTI)

Very high temperature gas reactors are helium-cooled, graphite-moderated advanced reactors that show potential for generating low-cost electricity via gas turbines or cogeneration with process-heat applications. This investigation addresses the development status of advanced coatings for nuclear-fuel particles and high-temperature structural materials and evaluates whether these developments are likely to lead to economically competitive applications of the very high temperature gas reactor concept.

1991-08-08T23:59:59.000Z

42

A SODIUM-GRAPHITE REACTOR STEAM-ELECTRIC STATION FOR 75 MEGAWATTS NET GENERATION  

SciTech Connect

The major design features, nuclear characteristics and performance data for a nuclear fueled central station power plant of 75,000 kw net capacity are presented. The heat source is a Na cooled graphite moderated reactor. The design of the reactor takes full advantage of the experience gained to date on the Sodium Reactor Experiment (SRE); the plant described here is a straightforward extension of the smaller experimental SRE, which is now under construction. The fuel elements are made up of rod clusters and the moderator is in the form of Zr canned graphite elements. The performance of the reactor has been based on conservative temperatures and coolant flow velocities which result in a plant with "built-in reserve." Thus, as experience is gained and anticipated improvements in reactor fuel elements and construction materials are proven, the performance of the plant can be increased accordingly. Two reactor designs are described, one for operation with slightly enriched U fuel elements and the other for operation with Th--U fuel elements. The associated heat exchangers, pumps, steam, and electrical generating equipment are identical for either reactor design. An analysis of turbine cycles describes the particular cycle chosen for initial operation and discusses a method by which modern central station performance can be initially obtained. The design and performance data which are required to enable reliable estimates of the plant construction and operating costs to be made are established. (auth)

Weisner, E.F.; Sybert, W.M.

1955-03-22T23:59:59.000Z

43

Advanced Reactor Research and Development Funding Opportunity...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Advanced Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy...

44

A SODIUM COOLED, GRAPHITE MODERATED, LOW ENRICHMENT URANIUM REACTOR FOR THE PRODUCTION OF USEFUL POWER  

SciTech Connect

A design study is presented for a sodium cooled, graphite moderated power reactor utilizing low enrichment uranium fuel. The design is characterized by dependence on existing technology and the use of standard, or nearly standard, components. The reactor has a nominal rating of 167 thermal megawatts, and a plant comprising three such reactors for a total output of 500 thermal megawatts is described. Sodium in a secondary, non-radioactive, circulation system carries the heat to a steam generator at 910 deg F and is returned at 420 deg F. Steam conditions at the turbine throttle are 600 psig and 825 deg F. Cost of the complete reactor power plant, consisting of the three reactors and one 150- megawatt turbogenerator, is estimated to be approximately ,165,000. (auth)

Weisner, E.F. ed.

1954-09-15T23:59:59.000Z

45

Present Status Of Research Reactor Decommissioning Program In Indonesia  

E-Print Network (OSTI)

At present, Indonesia has 3 research reactors: MTR-type multipurpose reactor of 30 MW at Serpong site, TRIGA-type research reactor of 1 MW at Bandung site, and small TRIGA - type reactor of 100 kW at Yogyakarta Research Center. The oldest one is the TRIGA reactor at Bandung site, which went critical at 250 kW in 1964, then was operated at maximum of 1000 kW by 1971. The reactor has operated for a total of 35 years. There is no decision for decommissioning this reactor; however, slowly but surely, it will be an object for a near-future decommissioning program. Anticipation of the situation is necessary. For the Indonesian case, early decommissioning strategy for a research reactor and restricted use of the site for another nuclear installation is favorable under high land pricing, availability of radwaste repository, and cost analysis. Graphite from Triga reactor reflector is recommended for direct disposal after conditioning, without volume reduction treatment. Development of human ...

Mulyanto And Gunandjar

2000-01-01T23:59:59.000Z

46

Variation of the magnetic susceptibility of artificial graphite with exposure in the materials testing reactor  

SciTech Connect

The magnetic susceptibility of artificial graphite was determined as a function of exposure in the MTR. Specimens were studied with exposures ranging from 0.07 to 82 {times} 10{sup18} nvt. Fluxes were determined by means of x-ray measurements and resistivity measurements. The dependence of the magnetic susceptibility on exposure in the MTR and also in a Hanford reactor are graphed, and an equivalence factor is calculated.

McCelland, J.D.

1955-02-23T23:59:59.000Z

47

PRELIMINARY DESIGN STUDY FOR A SODIUM-GRAPHITE-REACTOR IRRADIATION FACILITY  

SciTech Connect

The results of an investigation to integrate a Na/sup 24/ irradiation processing facility with an operating sodium graphite reactor are presented. An irradiation facility incorporated into a reference SGR (Hallam Nuclear Power Facility, Hallam, Nebraska) is described. Development of the facility application, preliminary design criteria and capital and operating costs are discussed. Recommendations for further development of the technology and economics of this type of irradiation facility are included. (auth)

Thompson, D.S.; Benaroya, V.

1959-01-31T23:59:59.000Z

48

FUEL CYCLE COSTS IN A GRAPHITE MODERATED SLIGHTLY ENRICHED FUSED SALT REACTOR  

SciTech Connect

A fuel cycle economic study has been made for a 315Mwe graphite- moderated slightly enriched fused-salt reactor. Fuel cycle costs of less than 1.5 mills may be possible for such reactors operating on a ten-year cycle even when the fuel is discarded at the end of the cycle. Recovery of the uranium and plutonium at the end of the cycle reduces the fuel cycle costs to approximates 1 mill/kwh. Changes in the waste storage cost, reprocessing cost or salt inventory have a relatively minor effect on fuel cycle costs. (auth)

Guthrie, C.E.

1959-01-01T23:59:59.000Z

49

2012 Annual Report Research Reactor Infrastructure Program  

SciTech Connect

The content of this report is the 2012 Annual Report for the Research Reactor Infrastructure Program.

Douglas Morrell

2012-11-01T23:59:59.000Z

50

Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics  

E-Print Network (OSTI)

Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermal neutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermal neutron diffusion length in graphite is evaluated at $50.85 \\pm 0.3cm$ using COMSOL Multiphysics and $50.95 \\pm 0.5cm$ using MCNP. Finally the total neutron cross-section is derived using FEA in an inverse iteration form.

S. R. Mirfayzi

2013-01-08T23:59:59.000Z

51

International Research Reactor Decommissioning Project  

SciTech Connect

Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement to the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.

Leopando, Leonardo [Philippine Nuclear Research Institute, Quezon City (Philippines); Warnecke, Ernst [International Atomic Energy Agency, Vienna (Austria)

2008-01-15T23:59:59.000Z

52

Reactor sharing experience at the MIT research reactor  

SciTech Connect

This paper provides a number of examples of how educational institutions in the Boston area and elsewhere that do not possess nuclear reactors for training and research purposes have successfully enriched their programs through utilization of the Massachusetts Institute of Technology Research Reactor (MITR) with assistance from the Reactor Sharing Program of the US Department of Energy (DOE).

Clark, L. Jr.; Fecych, W.; Young, H.H.

1985-01-01T23:59:59.000Z

53

Graphite Modular Reactor with Cooled Metal Core Outlet End Support Plate  

SciTech Connect

The modular designs appear attractive in that the reactor core lateral support is provided by the modules themselves rather than externally as with a bundled core. Types B and C provide a means of reducing the inter-element and intermodular leakage flow. This tends to enhance the reliability of the reactor core by decreasing the probability of a progressive overall failure of the core initiated by a mid-core rupture of one of the fuel elements. The graphite inner reflector and lateral support mechnism are eliminated in this design. The assembly of the modular core design is probably simplified by the smaller number of modular elements to be handled and by the elimination of the lateral support mechanism. The modular core has several potential problem areas which will be examined by further analysis and testing.

Jackson, L.

1963-08-01T23:59:59.000Z

54

Graphite Decommissioning  

Science Conference Proceedings (OSTI)

Many of the international participants in the EPRI Decommissioning Technology Program use graphite as a moderator material in their gas cooled reactors. This report reviews the current options for the management and disposal of irradiated nuclear graphite following the decommissioning of these nuclear installations. It also discusses specific issues associated with the disposal of graphite, and outlines innovative options for recycling or reusing products formed from the irradiated material.

2006-03-03T23:59:59.000Z

55

Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor  

SciTech Connect

The United States Department of Energys Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energys lead laboratory for nuclear energy development. The ATR is one of the worlds premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The final design phase for the first experiment was completed in September 2008, and the fabrication and assembly of the experiment test train as well as installation and testing of the control and support systems that will monitor and control the experiment during irradiation are being completed in early calendar 2009. The first experiment is scheduled to be ready for insertion in the ATR by April 30, 2009. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and data collection systems.

S. Blaine Grover

2009-05-01T23:59:59.000Z

56

Graphite Dust Deflagration  

Science Conference Proceedings (OSTI)

The graphite moderators of retired gas-cooled nuclear reactors present a difficult challenge during demolition activities. As part of the EPRI graphite initiative on the technical issues involved in the management and disposal of irradiated nuclear graphite, this report examines the international data on dust deflagration relevant to the decommissioning of graphite-moderated reactors. The report concludes that the risk of an explosion involving graphite dust during decommissioning is extremely low, and s...

2007-03-27T23:59:59.000Z

57

Graphite Dust Deflagration  

Science Conference Proceedings (OSTI)

The graphite moderators of retired gas-cooled nuclear reactors present a difficult challenge during demolition activities. As part of the EPRI graphite initiative on the technical issues involved in the management and disposal of irradiated nuclear graphite, EPRI Report 1014797 Graphite Dust Deflagration: A Review of International Data with Particular Reference to the Decommissioning of Graphite Moderated Reactors (March 2007) examined the international data on dust deflagration relevant to the decommiss...

2007-10-03T23:59:59.000Z

58

Decommssioning Of Research Reactor: Problemsand Experience  

E-Print Network (OSTI)

The study of the preparation for decommissioning the Research Reactor in Salaspils (Latvia) and the experience of decommissioning the Research Reactor in Sosny (Belarus) show that the problem of decommissioning research reactors is acute for countries that have no NPPs or their own nuclear industry. It also is associated with regulatory framework, planning and design, dismantling technologies, decontamination of radioactive equipment and materials, spent fuel and radioactive waste management, etc. 1. INTRODUCTION According to the IAEA research reactor database, there are about 300 research reactors worldwide [1]. At present over 30% of them have lifetimes of more than 35 years, 60% of more than 25 years. After the Chernobyl accident, significant efforts were made by many countries to modernize old research reactors aiming, first of all, at ensuring safe operation. However, a large number of aging research reactors will be facing shutdown in the near future. The problem of decommis...

Alexander Mikhalevich

2000-01-01T23:59:59.000Z

59

Status of the NGNP Graphite Creep Experiments AGC-1 and AGC-2 Irradiated in the Advanced Test Reactor  

Science Conference Proceedings (OSTI)

The United States Department of Energys Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.

Blaine Grover

2012-10-01T23:59:59.000Z

60

Research Reactors Division | Neutron Science | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

is responsible for operation of the High Flux Isotope Reactor. Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for research in the United States,...

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Reactor Safety Research Programs Quarterly Report October - December 1980  

SciTech Connect

This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S K

1981-04-01T23:59:59.000Z

62

THE NEXT GENERATION NUCLEAR PLANT GRAPHITE PROGRAM  

Science Conference Proceedings (OSTI)

Developing new nuclear grades of graphite used in the core of a High Temperature Gas-cooled Reactor (HTGR) is one of the critical development activities being pursued within the Next Generation Nuclear Plant (NGNP) program. Graphites thermal stability (in an inert gas environment), high compressive strength, fabricability, and cost effective price make it an ideal core structural material for the HTGR reactor design. While the general characteristics necessary for producing nuclear grade graphite are understood, historical nuclear grades no longer exist. New grades must be fabricated, characterized, and irradiated to demonstrate that current grades of graphite exhibit acceptable non-irradiated and irradiated properties upon which the thermo-mechanical design of the structural graphite in NGNP is based. The NGNP graphite R&D program has selected a handful of commercially available types for research and development activities necessary to qualify this nuclear grade graphite for use within the NGNP reactor. These activities fall within five primary areas; 1) material property characterization, 2) irradiated material property characterization, 3) modeling, and 4) ASTM test development, and 5) ASME code development efforts. Individual research and development activities within each area are being pursued with the ultimate goal of obtaining a commercial operating license for the nuclear graphite from the US NRC.

William E. Windes; Timothy D. Burchell; Robert L. Bratton

2008-09-01T23:59:59.000Z

63

FOREIGN RESEARCH AND POWER REACTOR PRELIMINARY LIST  

SciTech Connect

Foreign research and power reactors are tabulated. Nuclear power buildup goals are given for each nation on which information is available. (J.H.D.)

Ullmann, J.W.

1959-02-26T23:59:59.000Z

64

University Research Reactor Task Force to the Nuclear Energy Research  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

University Research Reactor Task Force to the Nuclear Energy University Research Reactor Task Force to the Nuclear Energy Research Advisory Committee University Research Reactor Task Force to the Nuclear Energy Research Advisory Committee In mid-February, 2001 The University Research Reactor (URR) Task Force (TF), a sub-group of the Department of Energy (DOE) Nuclear Energy Research Advisory Committee (NERAC), was asked to: * Analyze information collected by DOE, the NERAC "Blue Ribbon Panel," universities, and other sources pertaining to university reactors including their research and training capabilities, costs to operate, and operating data, and * Provide DOE with clear, near-term recommendations as to actions that should be taken by the Federal Government and a long-term strategy to assure the continued operation of vital university reactor facilities in

65

Estimation procedures and error analysis for inferring the total plutonium (Pu) produced by a graphite-moderated reactor  

Science Conference Proceedings (OSTI)

Graphite isotope ratio method (GIRM) is a technique that uses measurements and computer models to estimate total plutonium (Pu) production in a graphite-moderated reactor. First, isotopic ratios of trace elements in extracted graphite samples from the target reactor are measured. Then, computer models of the reactor relate those ratios to Pu production. Because Pu is controlled under nonproliferation agreements, an estimate of total Pu production is often required, and a declaration of total Pu might need to be verified through GIRM. In some cases, reactor information (such as core dimensions, coolant details, and operating history) are so well documented that computer models can predict total Pu production without the need for measurements. However, in most cases, reactor information is imperfectly known, so a measurement and model-based method such as GIRM is essential. Here, we focus on GIRMs estimation procedure and its associated uncertainty. We illustrate a simulation strategy for a specific reactor that estimates GIRMs uncertainty and determines which inputs contribute most to GIRMs uncertainty, including inputs to the computer models. These models include a local code that relates isotopic ratios to the local Pu production, and a global code that predicts the Pu production shape over the entire reactor. This predicted shape is included with other 3D basis functions to provide a hybrid basis set that is used to fit the local Pu production estimates. The fitted shape can then be integrated over the entire reactor to estimate total Pu production. This GIRM evaluation provides a good example of several techniques of uncertainty analysis and introduces new reasons to fit a function using basis functions in the evaluation of the impact of uncertainty in the true 3D shape.

Heasler, Patrick G.; Burr, Tom; Reid, Bruce D.; Gesh, Christopher J.; Bayne, Charles K.

2006-10-01T23:59:59.000Z

66

SAFEGUARD REPORT ON THE PROPOSED METHOD OF ANNEALING GRAPHITE IN THE X-10 REACTOR  

SciTech Connect

gone approximately 16 years of almost continuous irradiation. Throughout this time stored energy has accumulated at a slow rate to the present maximum value of about 35 cal/gm releasable to 250 deg C. A small portion of the moderator (approximately 4%) contains stored energy which under adiabatic conditions may be released spontaneously (at approximately 165 deg C) to produce a maximum temperature of 270 deg C. Careful analysis has shown that the presert condition is not hazardous; however, it appears wise at this time to initiate some corrective action (thermal annealing) to prevent the continued buildup of stored energy to a dangerously high value. Several methode of obtaining effective annealing in the OGR were investigated. The proposed method was selected upon the basis of convenience, over-all safety, effectiveness, and cost. The proposed method involves the alteration of the present coolant flow system to permit reversal of air flow through the fuel channels. This will result in a reversed temperature distribution wherein the maximum graphite temperature will occur in the normally cold, maximum-stored-energy region of the moderator, Such an arrangement permits an annealing operation to be performed under conditions very similar to those of the normal safe operation. The proposed procedure re quires a slow heatup of the moderator under full reversed, air-flow conditions. This can be accomplished by slowly raising the reactor nuclear power level until the desired graphite temperature is attained. During the initial stages of the operation the stored energy will be reduced to a sufficiently low value that spontaneous energy release is no longer possible. This can be accomplished at temperatures (less than 140 deg C) well below the minimum (165 deg C) required for spontaneous energy release. Subsequent higher temperature may then be employed to further reduce the stored energy. Recurrence of stored energy which may be released spontaneously can be prevented by periodic annealing at rather infrequent intervals. (auth)

Stanford, L.E.; Wittels, M.C.; Ramsey, M.E.; Cagle, C.D.

1960-05-18T23:59:59.000Z

67

Baseline Graphite Characterization: First Billet  

SciTech Connect

The Next Generation Nuclear Plant Project Graphite Research and Development program is currently establishing the safe operating envelope of graphite core components for a very high temperature reactor design. To meet this goal, the program is generating the extensive amount of quantitative data necessary for predicting the behavior and operating performance of the available nuclear graphite grades. In order determine the in-service behavior of the graphite for the latest proposed designs, two main programs are underway. The first, the Advanced Graphite Creep (AGC) program, is a set of experiments that are designed to evaluate the irradiated properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences, and compressive loads. Despite the aggressive experimental matrix that comprises the set of AGC test runs, a limited amount of data can be generated based upon the availability of space within the Advanced Test Reactor and the geometric constraints placed on the AGC specimens that will be inserted. In order to supplement the AGC data set, the Baseline Graphite Characterization program will endeavor to provide supplemental data that will characterize the inherent property variability in nuclear-grade graphite without the testing constraints of the AGC program. This variability in properties is a natural artifact of graphite due to the geologic raw materials that are utilized in its production. This variability will be quantified not only within a single billet of as-produced graphite, but also from billets within a single lot, billets from different lots of the same grade, and across different billets of the numerous grades of nuclear graphite that are presently available. The thorough understanding of this variability will provide added detail to the irradiated property data, and provide a more thorough understanding of the behavior of graphite that will be used in reactor design and licensing. This report covers the development of the Baseline Graphite Characterization program from a testing and data collection standpoint through the completion of characterization on the first billet of nuclear-grade graphite. This data set is the starting point for all future evaluations and comparisons of material properties.

Mark C. Carroll; Joe Lords; David Rohrbaugh

2010-09-01T23:59:59.000Z

68

2012 CHEMISTRY & PHYSICS OF GRAPHITIC CARBON MATERIALS GORDON RESEARCH CONFERENCE, JUNE 17-22, 2012  

Science Conference Proceedings (OSTI)

This conference will highlight the urgency for research on graphitic carbon materials and gather scientists in physics, chemistry, and engineering to tackle the challenges in this field. The conference will focus on scalable synthesis, characterization, novel physical and electronic properties, structure-properties relationship studies, and new applications of the carbon materials. Contributors

Fertig, Herbert

2012-06-22T23:59:59.000Z

69

Reactor Decommissioning Projects | Brookhaven National Laboratory  

NLE Websites -- All DOE Office Websites (Extended Search)

Brookhaven Graphite Research Reactor(BGRR) BGRR Overview BGRR Complex Description Decommissioning Decision BGRR Complex Cleanup Actions BGRR Documents BGRR Science &...

70

United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support  

SciTech Connect

The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

Douglas Morrell

2011-03-01T23:59:59.000Z

71

SODIUM GRAPHITE REACTOR QUARTERLY PROGRESS REPORT FOR OCTOBER-DECEMBER 1955. SECTION A. SECTION B  

DOE Green Energy (OSTI)

An analysis was made of the nuclear parameters for sodium graphite reactor lattices. These parameters include thermal utilization, macroscopic cross sections, thermal diffusion length, and neutron absorption. Results of all calculations are given in graphical form. Test fuel slugs for the SRE were cycled up to 500 times between 100 and 500 deg C at the rate of 2 cycles/hr. Results are tabulated. The centrifugal casting of U alloy fuel slugs is briefly evaluated. Results of the microscopic examination of the extruded ThU breeder fuels are shown. The percent elongation of graphite due to the presence of Na is shown for various temperatures. Results of wear tests on graphite are also tabulated. The behavior of Zr in liquid Na was studied, and weight gains in Zr are summarized. Analog computer studies were continued, and data are included on the temperature effects of the response time of coolant channel Na outlet temperature thermocouples, the effects of continuous rod motion and pump speed changes on the outlet Na temperature and power, and the outlet temperature as a function of scram time. The critical evaluation of B--Ni rods is tabulated. The fuel rod assembly apparatus is described. Fuel rod development is discussed. Cyclograph traces of rods bonded with various Na--K alloys were recorded for rods at room temperature and heated to 450 and 600 deg F. The traces are indicative of uniform bonding. The moderator can fabrication and testing is also discussed. Tests were completed on Freeze Seal No. 2 for the 6-in. oval port Wedgeplug test valve at 450, 850, and 1250 deg F. The temperature gradients from the hot flange face to the end of the seal mechanism for various valve temperature conditions are shown. Sodium leak rates through the valve are tabulated. Progress in the development of a liquid Na level gage is briefly reported. The tubular heater experiment was completed, and the times to raise pipe temperatures from ambient to 350 deg F are tabulated. Designs for a 6-in. Na pump loop are described briefly. A one to 3.5 scale model of a SRE fuel element was constructed to study the effect of side drag on the element during insertion operations at those fuel channels located near the outlet of the upper plenum chamber. The calibration of the SRE fuel element orifices was studied. Control rod lead screw development is discussed. Development of the safety rod system is described. Core tank galling tests are summarized. Experiments to determine the effects of radiation on MoS/ sub 2/ are described. Dose buildup factors for the concretes to be used in the reactor top shield are tabulated. Constants for the quadratic representation of the dose buildup factor and the capture gamma rays from heavy concrete are also tabulated. The sodium pumps and service system are described. The rated cooling capacity of the SRE tetralin evaporative cooler was checked. Results of a study on the effects of NaK temperature on the H content of He are tabulated. (D.E.B.)

Martin, A.B.; Cochran, J.C. ed.

1956-04-15T23:59:59.000Z

72

Relicensing of the MIT Research Reactor  

SciTech Connect

The Massachusetts Institute of Technology (MIT) Research Reactor (MITR) is owned and operated by MIT, a nonprofit university. The current reactor, MITR-II, is a 5-MW, light water-cooled and heavy water-moderated reactor that uses materials test reactor-type fuel. Documents supporting application to the U.S. Nuclear Regulatory Commission (NRC) for relicensing of MITR were submitted in July 1999. A power upgrade from 5 to 6 MW was also requested. The relicensed reactor (MITR-III) will be the third reactor operated by MIT. This paper describes MITR-I and MITR-II, and design options considered for MITR-III. Selected problems addressed during the relicensing studies are also described, namely core tank aging evaluation, neutronic analysis, thermal-hydraulic analysis, and step reactivity insertion analysis.

Lin-Wen Hu; John A. Bernard; Susan Tucker

2000-06-04T23:59:59.000Z

73

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report  

SciTech Connect

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed.

Hauptman, H.M.; Petro, J.N.; Jacobi, O. [and others

1995-04-01T23:59:59.000Z

74

Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel  

DOE Green Energy (OSTI)

The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [{approx}1000 Mw(e)] with passive safety systems to provide the potential for improved economics.

Forsberg, C.W.

2002-02-21T23:59:59.000Z

75

License extension for the MIT research reactor  

Science Conference Proceedings (OSTI)

On February 8, 1995, the U.S. Nuclear Regulatory Commission (NRC) amended the operating license for the Massachusetts Institute of Technology (MIT) research reactor (MITR) by extending its expiration date from May 7, 1996, to August 8, 1999. This increase of 2.25 yr in the duration of the license was for the purpose of recapturing construction time. Extensions of this type have routinely been granted to power plants. The significance of the present action is that it is the first such extension ever granted to a research reactor. This paper describes the basis for the amendment request and the supporting safety analysis.

Bernard, J.A. [Massachusetts Institute of Technology, Cambridge, MA (United States)

1995-12-31T23:59:59.000Z

76

Brazing graphite to graphite  

DOE Patents (OSTI)

Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of virtually graphite.

Peterson, George R. (Andersonville, TN)

1976-01-01T23:59:59.000Z

77

Safer nuclear reactors could result from Los Alamos research  

NLE Websites -- All DOE Office Websites (Extended Search)

Calendar Video Newsroom News Releases News Releases - 2010 March Safer nuclear reactors could result from research Safer nuclear reactors could result from Los...

78

The triple axis spectrometer at the new research reactor OPAL ...  

Science Conference Proceedings (OSTI)

... The triple axis spectrometer at the new research reactor OPAL in Australia. ... The TAS will be based on a thermal beam at the reactor face. ...

79

Evaluation of graphite/steam interactions for ITER (International Thermonuclear Experimental Reactor)  

DOE Green Energy (OSTI)

In this report we present the results of an experimental/analytical study designed to determine the quantity of hydrogen generated during a coolant inleakage accident in ITER. This hydrogen could represent a potential explosive hazard, provided the proper conditions exist, causing machine damage and release of radioactive material. We have measured graphite/steam reaction rates for several graphites and carbon-based composites at temperatures between 1000 C and 1700 C. The effects of steam flow rate, and partial pressure were also examined. The measured reaction rates correlated well with two Arrhenius type relationships. We have used the relationships for GraphNOL N3M in a thermal model to determine that for ITER the quantity of hydrogen produced would range between 5 and 35 kg, depending upon how the graphite tiles are attached to the first wall. While 5 kg is not a significant concern, 35 kg presents an explosive hazard. 20 refs., 14 figs., 1 tab.

Smolik, G.R.; Merrill, B.J.; Piet, S.J.; Holland, D.F.

1990-09-01T23:59:59.000Z

80

Research and Medical Isotope Reactor Supply | Y-12 National Security...  

NLE Websites -- All DOE Office Websites (Extended Search)

Research and Medical ... Research and Medical Isotope Reactor Supply Our goal is to fuel research and test reactors with low-enriched uranium. Y-12 tops the short list of the...

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Comprehensive Thermal Hydraulics Research of the Very High Temperature Gas Cooled Reactor  

SciTech Connect

The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; David Petti; Hyung Kang

2010-10-01T23:59:59.000Z

82

Baseline Graphite Initial Mechanical Test Report  

SciTech Connect

The Next Generation Nuclear Plant (NGNP) Project is tasked with selecting a high temperature gas reactor technology that will be capable of generating electricity and supplying large amounts of process heat. The NGNP is presently being designed as a helium-cooled high temperature gas reactor (HTGR) with a large graphite core. The graphite baseline characterization project is conducting the research and development (R&D) activities deemed necessary to fully qualify nuclear-grade graphite for use in the NGNP reactor. One of the major fundamental objectives of the project is establishing nonirradiated thermomechanical and thermophysical properties by characterizing lot-to-lot and billet-to-billet variations (for probabilistic baseline data needs) through extensive data collection and statistical analysis. The reactor core will be made up of stacks of graphite moderator blocks. In order to gain a more comprehensive understanding of the varying characteristics in a wide range of suitable graphites, any of which can be classified as nuclear grade, an experimental program has been initiated to develop an extensive database of the baseline characteristics of numerous candidate graphites. Various factors known to affect the properties of graphite will be investigated, including specimen size, spatial location within a graphite billet, specimen orientation within a billet (either parallel to [P] or transverse to [T] the long axis of the as-produced billet), and billet-to-billet variations within a lot or across different production lots. Because each data point is based on a certain position within a given billet of graphite, particular attention must be paid to the traceability of each specimen and its spatial location and orientation within each billet. The evaluation of these properties is discussed in the Graphite Technology Development Plan (Windes et. al 2007). One of the key components in the evaluation of these graphite types will be mechanical testing of specimens drawn from carefully controlled sections of each billet. This report is confirmation that the test procedures are in place and approved, and that mechanical testing of graphite under the Baseline Graphite Characterization program has commenced.

Mark Carroll; Randy Lloyd

2009-09-01T23:59:59.000Z

83

Ames Laboratory Research Reactor Facility Ames, Iowa  

Office of Legacy Management (LM)

,, *' ; . Final Radiological Condition of the Ames Laboratory Research Reactor Facility Ames, Iowa _, . AGENCY: Office of Operational Safety, Department of Energy ' ACTION: Notice of Availability of Archival Information Package SUMMARY: The'Office of Operational Safety of the Department O i Energy (DOE) has reviewed documentation relating to the decontamination and decommissioning operations conducted at the Ames Laboratory Research Reactor Facility, Ames, Iowa and has prepared an archival informati0.n package to permanently document the results of the action and the site conditions and use restriction placed on the . site at the tim e of release. This review is based on post-decontamination survey data and other pertinent documentation referenced in and included in the archival package. The material and

84

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980  

SciTech Connect

Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

Not Available

1981-08-01T23:59:59.000Z

85

Sterile Neutrino Search Using China Advanced Research Reactor  

E-Print Network (OSTI)

We study the feasibility of a sterile neutrino search at the China Advanced Research Reactor by measuring $\\bar {\

Guo, Gang; Ji, Xiangdong; Liu, Jianglai; Xi, Zhaoxu; Zhang, Huanqiao

2013-01-01T23:59:59.000Z

86

Reactor operations: Brookhaven Medical Research Reactor, Brookhaven High Flux Beam Reactor. Informal report, June 1995  

Science Conference Proceedings (OSTI)

Part one of this report gives the operating history of the Brookhaven Medical Research Reactor for the month of June. Also included are the BMRR technical safety surveillance requirements record and the summary of BMRR irradiations for the month. Part two gives the operating histories of the Brookhaven High Flux Beam Reactor and the Cold Neutron Facility at HFBR for June. Also included are the HFBR technical safety surveillance requirements record and the summary of HFBR irradiations for the month.

NONE

1995-06-01T23:59:59.000Z

87

Reactor operations: Brookhaven Medical Research Reactor, Brookhaven High Flux Beam Reactor. Informal report, July 1995  

Science Conference Proceedings (OSTI)

Part one of this report gives the operating history for the Brookhaven Medical Research Reactor for the month of July. Also included are the BMRR technical safety surveillance requirements record and the summary of BMRR irradiations for the month. Part two gives the operating histories for the Brookhaven High Flux Beam Reactor and the Cold Neutron Source Facility for the month of July. Also included are the HFBR technical safety surveillance requirements record and the summary of HFBR irradiations for the month.

NONE

1995-07-01T23:59:59.000Z

88

The Argonaut Reactor - Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Argonne Reactors > Training Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

89

Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program  

DOE Green Energy (OSTI)

Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions.

Not Available

1987-05-01T23:59:59.000Z

90

Research Reactors Division | ORNL Neutron Sciences  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactors Division (RRD) is responsible for operation of the High Flux Isotope Reactor (HFIR). Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for...

91

Global estimation of potential unreported plutonium in thermal research reactors  

SciTech Connect

As of November, 1993, 303 research reactors (research, test, training, prototype, and electricity producing) were operational worldwide; 155 of these were in non-nuclear weapon states. Of these 155 research reactors, 80 are thermal reactors that have a power rating of 1 MW(th) or greater and could be utilized to produce plutonium. A previously published study on the unreported plutonium production of six research reactors indicates that a minimum reactor power of 40 MW (th) is required to make a significant quantity (SQ), 8 kg, of fissile plutonium per year by unreported irradiations. As part of the Global Nuclear Material Control Model effort, we determined an upper bound on the maximum possible quantity of plutonium that could be produced by the 80 thermal research reactors in the non-nuclear weapon states (NNWS). We estimate that in one year a maximum of roughly one quarter of a metric ton (250 kg) of plutonium could be produced in these 80 NNWS thermal research reactors based on their reported power output. We have calculated the quantity of plutonium and the number of years that would be required to produce an SQ of plutonium in the 80 thermal research reactors and aggregated by NNWS. A safeguards approach for multiple thermal research reactors that can produce less than 1 SQ per year should be conducted in association with further developing a safeguards and design information reverification approach for states that have multiple research reactors.

Dreicer, J.S.; Rutherford, D.A.

1996-09-01T23:59:59.000Z

92

THE HGCR-1, A DESIGN STUDY OF A NUCLEAR POWER STATION EMPLOYING A HIGH- TEMPERATURE GAS-COOLED REACTOR WITH GRAPHITE-UO$sub 2$ FUEL ELEMENTS  

SciTech Connect

The preliminary design of a 3095-Mw(thermal), helium-cooled, graphite- moderated reactor employing sign conditions, 1500 deg F reactor outlet gas would be circulated to eight steam generators to produce 1050 deg F, 1450-psi steam which would be converted to electrical power in eight 157-Mw(electrical) turbine- generators. The over-all efficiency of this nuclear power station is 36.5%. The significant activities released from the unclad graphite-UO/sub 2/ fuel appear to be less than 0.2% of those produced and would be equivalent to 0.002 curie/ cm/ sup 3/ in the primary helium circuit. The maintenance problems associated with this contamination level are discussed. A cost analysis indicates that the capital cost of this nuclear station per electrical kilowatt would be around 0, and that the production cost of electrical power would be 7.8 mills/kwhr. (auth)

Cottrell, W.B.; Copenhaver, C.M.; Culver, H.N.; Fontana, M.H.; Kelleghan, V.J.; Samuels, G.

1959-07-28T23:59:59.000Z

93

The LEU conversion status of U.S. Research Reactors.  

SciTech Connect

This paper summarizes the conversion status of US research and test reactors and estimates uranium densities needed to convert reactors with power levels 21 MW from HEU ({ge} 20% U-235) to LEU (<20% U-235) fuels. Detailed conversion studies for each reactor need to be completed in order to establish the feasibility of using LEU fuels.

Matos, J. E.

1997-11-14T23:59:59.000Z

94

Progressive Application Decommissioning Models for U.S. Power and Research Reactors  

SciTech Connect

This paper presents progressive engineering techniques and experiences in decommissioning projects performed by Bums and Roe Enterprises within the last fifteen years. Specifically, engineering decommissioning technical methods and lessons learned are discussed related to the Trojan Large Component Removal Project, San Onofre Nuclear Generating Station (SONGS) Decommissioning Project and the Brookhaven Graphite Research Reactor (BGRR) Decommissioning Project Study. The 25 years since the 1979 TMI accident and the events following 9/11 have driven the nuclear industry away from excessive, closed/elitist conservative methods towards more pragmatic results-oriented and open processes. This includes the essential recognition that codes, standards and regulatory procedures must be efficient, effective and fit for purpose. Financial and open-interactive stakeholder pressures also force adherence to aggressive risk reduction posture in the area of a safety, security and operations. The engineering methods and techniques applied to each project presented unique technical solutions. The decommissioning design for each project had to adopt existing design rules applicable to construction of new nuclear power plants and systems. It was found that the existing ASME, NRC, and DOE codes and regulations for deconstruction were, at best, limited or extremely conservative in their applicability to decommissioning. This paper also suggests some practical modification to design code rules in application for decommissioning and deconstruction. The representative decommissioning projects, Trojan, SONGS and Brookhaven, are discussed separately and the uniqueness of each project, in terms of engineering processes and individual deconstruction steps, is discussed. Trojan Decommissioning. The project included removal of entire NSSS system. The engineering complexity was mainly related to the 1200 MW Reactor. The approach, process of removal, engineering method related to protect the worker against excessive radiation exposure, transportation, and satisfying applicable rules and regulations, were the major problems to overcome. The project's successful completed earned a patent award. SONGS Decommissioning. The reactor's spherical containment and weakened integrity was the scope of this decommissioning effort. The aspects of structure stability and method of deconstruction is the major part of the presentation. The economical process of deconstruction, aspects of structural stability, worker safety, and the protection of the surrounding environment from contamination is highlighted in this section. BGRR Decommissioning Study. BREI was commissioned by Brookhaven National Laboratory (BNL) to evaluate and analyze the stability, and progressive decommissioning, and removal of BGRR components. This analysis took the form of several detailed decommissioning studies that range from disassembly and removal of the unit's graphite pile to the complete environmental restoration of the reactor site. While most of the facility's decommissioning effort is conventional, the graphite pile and its biological shield present the greatest challenge. The studies develop a unique method of removing high-activity waste trapped in the graphite joints. (authors)

Studnicka, Z.; Lacy, N.H.; Nicholas, R.G.; Campagna, M.; Morgan, R.D. [Bums and Roe Enterprises, Inc., 800 Kinderkamack Road, Oradell, NJ 07649 (United States); Sawruk, W. [ABS Consulting, Inc., 5 Birdsong Court, Shillington, PA 19607 (United States)

2006-07-01T23:59:59.000Z

95

N-Reactor Department Research and Development budget for FY 1966 and revision of budget for FY 1965  

SciTech Connect

The N-Reactor Department Research and Development Program for FY 1965, 1966, and later years is structured to achieve the following general goals. (1) Assurance of a high level of nuclear safety; (2) Assurance of achieving full plant life; (3) Reduction in operating costs for a given production rate; (4) Increase in production rate without proportionate increase in operating costs; (5) Savings in capital outlays necessary to achieve stated reductions in operating cost or increases in production; (6) Production of new products of value; (7) Savings in capital outlays or operating costs to achieve a given level of plant safety. The program is divided into three general categories; Reactor, Metallurgy, and Co-Product. The Reactor category is further divided into physics studies, thermal hydraulics studies, zircaloy process tube development, control, instrument and system analyses, chemistry, engineering research and development, gas, atmosphere studies, graphite studies, and nuclear safety research.

1964-03-25T23:59:59.000Z

96

Reactor Safety Research Programs Quarterly Report July - September 1981  

SciTech Connect

This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR} steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S. K.

1982-01-01T23:59:59.000Z

97

Computer-Based Model of the MIT Research Reactor  

SciTech Connect

A description is given of a model of the Massachusetts Institute of Technology Research Reactor (MITR) in which both the reactor's neutronic and thermal-hydraulic behaviors are replicated. The purpose of the model is to support control studies and the development of techniques for the automated diagnosis of reactivity transients. In particular, comparison of the model's predictions with actual measurements from the reactor will allow determination of whether the reactor is functioning as expected.

John A. Bernard; Lin-Wen Hu

2000-11-12T23:59:59.000Z

98

Opportunities for the Precision Study of Reactor Antineutrinos at Very Short Baselines at US Research Reactors  

E-Print Network (OSTI)

: pieter.mumm@nist.gov #12;2 past reactor experiments HFIR, ORNL NBSR, NIST ATR, INL available baselines at US research reactors 3 neutrino fit 3+1 neutrino fit Tuesday, August 7, 12 NIST ILL HFIR ATR SONGSNIST ILL HFIR ATR SONGS 10. 100 1000 core size reactor power reactorpower(MWth) 1meter ILL HFIR NBSR

99

FISSION PRODUCT TRAPS FOR USE IN HIGH-TEMPERATURE GAS-COOLED GRAPHITE REACTORS  

SciTech Connect

A proposal is given of an approach to a fission-product trapping system which appears feasible on the basis of thermodynamic and other data available. Reactor and trapping conditions are outlined. The half-lives, fission yields, and volatility of the fission products of interest are described. To provide the most effective retention at elevated temperatures, two types of reagents are required: a highly electropositive metal that will not melt or appreciably vaporize and which will form stable non-volatile compounds with non-metallic or near non-metallic fission products; and a reagent to provide a highly electronegative element to form stable, non-volatile compounds with metallic fission products. Thermodynamic properties are included for compounds formed by reactions between the fission products and the trapping reagents. (B.O.G.)

Zumwalt, L.R.

1958-03-13T23:59:59.000Z

100

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program The Department of Energy's (DOE's) Light Water Reactor Sustainability (LWRS) Program is a five year effort that works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operation of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear  

National Nuclear Security Administration (NNSA)

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Video Gallery > Maria Research Reactor loaded with LEU - ... Maria Research Reactor loaded with LEU - Otwock, Poland Maria Research Reactor loaded with LEU - Otwock, Poland

102

RERTR program reduces use of enriched uranium in research reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

RERTR program reduces use of enriched uranium in research reactors RERTR program reduces use of enriched uranium in research reactors worldwide Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share RERTR program reduces use of enriched uranium in research reactors worldwide The High Flux Reactor in Petten, the Netherlands READY TO CONVERT - The High Flux Reactor in Petten, the Netherlands, has

103

Program on Technology Innovation: Graphite Waste Separation  

Science Conference Proceedings (OSTI)

The graphite moderators of retired gas-cooled nuclear reactors present a difficult challenge during demolition activities. There is a widespread view that disposal would be greatly facilitated if carbon-14 could be removed from the graphite blocks. As part of the EPRI graphite initiative on the technical issues involved in the management and disposal of irradiated nuclear graphite, this report describes an engineering feasibility study of graphite radioisotope separation technology. The report evaluates ...

2008-03-10T23:59:59.000Z

104

Advanced Reactor Research and Development Funding Opportunity Announcement  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Research and Development Funding Opportunity Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) sponsors a program of research, development, and demonstration related to advanced non-light water reactor concepts. A goal of the program is to facilitate greater engagement between DOE and industry. During FY12, DOE established a Technical Review Panel (TRP) process to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. That process involved the use of a Request for Information (RFI) to solicit concept information from industry and engage technical experts to evaluate those concepts. Having completed this process, DOE desires to

105

Advanced Reactor Research and Development Funding Opportunity Announcement  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Advanced Reactor Research and Development Funding Opportunity Advanced Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) sponsors a program of research, development, and demonstration related to advanced non-light water reactor concepts. A goal of the program is to facilitate greater engagement between DOE and industry. During FY12, DOE established a Technical Review Panel (TRP) process to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. That process involved the use of a Request for Information (RFI) to solicit concept information from industry and engage technical experts to evaluate those concepts. Having completed this process, DOE desires to

106

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

SciTech Connect

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. Many 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.

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

2011-07-18T23:59:59.000Z

107

Sandia National Laboratories: Research: Facilities: Sandia Pulsed Reactor  

NLE Websites -- All DOE Office Websites (Extended Search)

Sandia Pulsed Reactor Facility - Critical Experiments Sandia Pulsed Reactor Facility - Critical Experiments Sandia scientist John Ford places fuel rods in the Seven Percent Critical Experiment (7uPCX) at the Sandia Pulsed Reactor Facility Critical Experiments (SPRF/CX) test reactor - a reactor stripped down to its simplest form. The Sandia Pulsed Reactor Facility - Critical Experiments (SPRF/CX) provides a flexible, shielded location for performing critical experiments that employ different reactor core configurations and fuel types. The facility is also available for hands-on nuclear criticality safety training. Research and other activities The 7% series, an evaluation of various core characteristics for higher commercial-fuel enrichment, is currently under way at the SPRF/CX. Past critical experiments at the SPRF/CX have included the Burnup Credit

108

Sodium fast reactor safety and licensing research plan. Volume II.  

SciTech Connect

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d'%C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

2012-05-01T23:59:59.000Z

109

Light Water Reactors Technology Development - Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

110

Engineering activities at the MIT research reactor in support of power reactor technology  

SciTech Connect

The Massachusetts Institute of Technology (MIT) research reactor (MITR-II) is a 5-MW(thermal) light-water-cooled and-moderated reactor (LWR) with in-core neutron and gamma dose rates that closely approximate those in current LWRs. Compact in-pile loops that simulate pressurized water reactor (PWR) and boiling water reactor (BWR) thermal hydraulics and coolant chemistry have been designed for installation in the MITR-II. A PWR loop has been completed and is currently operating in the reactor. A BWR loop is under construction, and an in-pile facility for irradiation-assisted stress corrosion crack (IASCC) testing is being designed. Another major area of research and on-line testing is the closed-loop, nonlinear, digital control of various reactor parameters, including the power level, temperature, and net energy production.

Harling, O.K.; Bernard, J.A.; Driscoll, M.J.; Kohse, G.E.; Ballinger, R.G.

1989-01-01T23:59:59.000Z

111

Automated startup of the MIT research reactor  

SciTech Connect

This summary describes the development, implementation, and testing of a generic method for performing automated startups of nuclear reactors described by space-independent kinetics under conditions of closed-loop digital control. The technique entails first obtaining a reliable estimate of the reactor's initial degree of subcriticality and then substituting that estimate into a model-based control law so as to permit a power increase from subcritical on a demanded trajectory. The estimation of subcriticality is accomplished by application of the perturbed reactivity method. The shutdown reactor is perturbed by the insertion of reactivity at a known rate. Observation of the resulting period permits determination of the initial degree of subcriticality. A major advantage to this method is that repeated estimates are obtained of the same quantity. Hence, statistical methods can be applied to improve the quality of the calculation.

Kwok, K.S. (Massachusetts Inst. of Tech., Cambridge (United States))

1992-01-01T23:59:59.000Z

112

Graphite design handbook  

Science Conference Proceedings (OSTI)

The objectives of the Graphite Design Handbook (GDH) are to provide and maintain a single source of graphite properties and phenomenological model of mechanical behavior to be used for design of MHTGR graphite components of the Reactor System, namely, core support, permanent side reflector, hexagonal reflector elements, and prismatic fuel elements; to provide a single source of data and material models for use in MHTGR graphite component design, performance, and safety analyses; to present properties and equations representing material models in a form which can be directly used by the designer or analyst without the need for interpretation and is compatible with analytical methods and structural criteria used in the MHTGR project, and to control the properties and material models used in the MHTGR design and analysis to proper Quality Assurance standards and project requirements. The reference graphite in the reactor internal components is the nuclear grade 2020. There are two subgrades of interest, the cylinder nuclear grade and the large rectangular nuclear grade. The large rectangular nuclear grade is molded in large rectangular blocks. It is the reference material for the permanent side reflector and the central column support structure. The cylindrical nuclear grade is isostatically pressed and is intended for use as the core support component. This report gives the design properties for both H-451 and 2020 graphite as they apply to their respective criteria. The properties are presented in a form for design, performance, and safety calculations that define or validate the component design. 103 refs., 20 figs., 19 tabs.

Ho, F.H.

1988-09-01T23:59:59.000Z

113

Uranium Oxide Aerosol Transport in Porous Graphite  

Science Conference Proceedings (OSTI)

The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactors lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

Blanchard, Jeremy; Gerlach, David C.; Scheele, Randall D.; Stewart, Mark L.; Reid, Bruce D.; Gauglitz, Phillip A.; Bagaasen, Larry M.; Brown, Charles C.; Iovin, Cristian; Delegard, Calvin H.; Zelenyuk, Alla; Buck, Edgar C.; Riley, Brian J.; Burns, Carolyn A.

2012-01-23T23:59:59.000Z

114

Argonne's pyroprocessing and advanced reactor research featured on WGN  

NLE Websites -- All DOE Office Websites (Extended Search)

Argonne's pyroprocessing and advanced reactor research featured on WGN Argonne's pyroprocessing and advanced reactor research featured on WGN radio Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share Argonne's pyroprocessing and advanced reactor research featured on WGN radio Uranium dendrites These tiny branches, or "dendrites," of pure uranium form when engineers

115

Nuclear reactor and materials science research: Final technical report, May 1, 1985-September 30, 1986. [Academic and research utilization of reactor  

SciTech Connect

Throughout the 17-month period of the grant, May 1, 1985 - September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The period encompassed MIT's fiscal year utilization of the reactor during that period may be classified as follows: neutron beam tube research, nuclear materials research and development, radiochemistry and trace analysis, nuclear medicine, radiation health physics, computer control of reactors, dose reduction in nuclear power reactors, reactor irradiations and services for groups outside MIT, and MIT research reactor. This paper provides detailed information on this research academic utilization.

Harling, O.K.

1987-05-11T23:59:59.000Z

116

Graphite technology development plan  

Science Conference Proceedings (OSTI)

This document presents the plan for the graphite technology development required to support the design of the 350 MW(t) Modular HTGR within the US National Gas-Cooled Reactor Program. Besides descriptions of the required technology development, cost estimates, and schedules, the plan also includes the associated design functions and design requirements.

NONE

1986-07-01T23:59:59.000Z

117

DISMANTLING OF THE REACTOR BLOCK OF THE FRJ-1 RESEARCH REACTOR (MERLIN)  

SciTech Connect

This report describes the past procedure in dismantling the reactor block of the FRJ-1 research reactor (MERLIN). Furthermore, it gives an outlook on future activities up to the final removal of the reactor block. MERLIN is an abbreviation for Medium Energy Research Light Water Moderated Industrial Nuclear Reactor. The FRJ-1 (MERLIN) was shut down in 1985 and the fuel elements removed from the facility. After dismantling the coolant loops and removing the reactor tank internals with subsequent draining of the reactor tank water, the first activities for dismantling the reactor block were carried out in summer 2001. The relevant license was granted in late July 2001 by the licensing authority specifying 8 incidental provisions. After dismantling the reactor extension (gates of the thermal columns and steel platforms surrounding the reactor block), a heavy-load platform including a casing around the reactor block was constructed. Two ventilation systems with a volume flow of 10,000 and 2 ,000 m3/h will, moreover, serve to avoid a spread of contamination. The reactor block will be dismantled in three phases divided according to upper, central and bottom sections. Dismantling the upper section started in August 2002. This section as well as the bottom section can probably be completely measured for clearance. For this reason, the activities have so far been carried out manually using mechanical and thermal techniques. The central section will probably have to be largely disposed of as radioactive waste. This is the region of the former reactor core in which the experimental devices are also integrated. Most of this work will probably have to be carried out by remote handling. More than 80 % of the dismantled materials of the reactor block can probably be measured for clearance. For this purpose, a clearance measurement device was taken into operation in the FRJ-1. On this occasion, the limits of clearance measurement have become evident. For concrete, which constitutes the largest portion of the dismantled materials by volume, an additional conditioning step has become necessary to fulfill the clearance criteria, whereas waste packages with steel components largely have to be reconditioned once more at a later stage. Material measured for clearance will be disposed of conventionally (recycling, landfill) after inspection by the official expert and clearance by the regulatory authority. Dismantled parts that cannot be measured for clearance will be transferred to the Decontamination Department of the Research Centre. From the present perspective, the dismantling of the reactor block will be completed within the first six months of 2003.

Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

2003-02-27T23:59:59.000Z

118

Spent graphite fuel element processing  

SciTech Connect

The Department of Energy currently sponsors two programs to demonstrate the processing of spent graphite fuel elements. General Atomic in San Diego operates a cold pilot plant to demonstrate the processing of both US and German high-temperature reactor fuel. Exxon Nuclear Idaho Company is demonstrating the processing of spent graphite fuel elements from Rover reactors operated for the Nuclear Rocket Propulsion Program. This work is done at Idaho National Engineering Laboratory, where a hot facility is being constructed to complete processing of the Rover fuel. This paper focuses on the graphite combustion process common to both programs.

Holder, N.D.; Olsen, C.W.

1981-07-01T23:59:59.000Z

119

LEU conversion status of US research reactors, September 1996  

SciTech Connect

This paper summarizes the conversion status of research and test reactors in the United States from the use of fuels containing highly- enriched uranium (HEU, greater than or equal to 20%) to the use of fuels containing low-enriched uranium (LEU, < 20%). Estimates of the uranium densities required for conversion are made for reactors with power levels greater than or equal to 1 MW that are not currently involved in the LEU conversion process.

Matos, J.E.

1996-10-07T23:59:59.000Z

120

Evaluation of co-cokes from bituminous coal with vacuum resid or decant oil, and evaluation of anthracites, as precursors to graphite.  

E-Print Network (OSTI)

??Graphite is utilized as a neutron moderator and structural component in some nuclear reactor designs. During the reactor operaction the structure of graphite is damaged (more)

Nyathi, Mhlwazi

2011-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Reduced enrichment for research and test reactors: Proceedings  

SciTech Connect

The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

1988-05-01T23:59:59.000Z

122

Research reactor usage at the Idaho National Engineering Laboratory in support of university research and education  

SciTech Connect

The Idaho National Engineering Laboratory is a US Department of Energy laboratory which has a substantial history of research and development in nuclear reactor technologies. There are a number of available nuclear reactor facilities which have been incorporated into the research and training needs of university nuclear engineering programs. This paper addresses the utilization of the Advanced Reactivity Measurement Facility (ARMF) and the Coupled Fast Reactivity Measurement Facility (CFRMF) for thesis and dissertation research in the PhD program in Nuclear Science and Engineering by the University of Idaho and Idaho State University. Other reactors at the INEL are also being used by various members of the academic community for thesis and dissertation research, as well as for research to advance the state of knowledge in innovative nuclear technologies, with the EBR-II facility playing an essential role in liquid metal breeder reactor research. 3 refs.

Woodall, D.M.; Dolan, T.J.; Stephens, A.G. (Idaho National Engineering Lab., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

123

Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

J. K. Wright; R. N. Wright

2010-07-01T23:59:59.000Z

124

Reactor Safety Research: Semiannual report, July-December 1986  

DOE Green Energy (OSTI)

Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions.

Not Available

1987-11-01T23:59:59.000Z

125

New Research Center to Increase Safety and Power Output of U.S. Nuclear Reactors  

Energy.gov (U.S. Department of Energy (DOE))

The Department of Energy dedicated the Consortium for Advanced Simulation of Light Water Reactors (CASL), an advanced research facility that will accelerate the advancement of nuclear reactor technology.

126

GRAPHITE EXTRUSIONS  

DOE Patents (OSTI)

A new lubricant for graphite extrusion is described. In the past, graphite extrusion mixtures have bcen composed of coke or carbon black, together with a carbonaceous binder such as coal tar pitch, and a lubricant such as petrolatum or a colloidal suspension of graphite in glycerin or oil. Sinee sueh a lubricant is not soluble in, or compatible with the biiider liquid, such mixtures were difficult to extrude, and thc formed pieees lacked strength. This patent teaches tbe use of fatty acids as graphite extrusion lubricants and definite improvemcnts are realized thereby since the fatty acids are soluble in the binder liquid.

Benziger, T.M.

1959-01-20T23:59:59.000Z

127

Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.

J. K. Wright; R. N. Wright

2008-04-01T23:59:59.000Z

128

RERTR 2009 (Reduced Enrichment for Research and Test Reactors)  

SciTech Connect

The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Test Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.

Totev, T.; Stevens, J.; Kim, Y. S.; Hofman, G.; Matos, J.; Hanan, N.; Garner, P.; Dionne, B.; Olson, A.; Feldman, E.; Dunn, F.; Nuclear Engineering Division; Atomic Research Center; Inst. of Nuclear Physics; LLNL; INL; Korea Atomic Energy Research Inst.; Comisi?n Nacional de Energ?a At?mica; Nuclear Reactor Lab.; Inst. of Atomic Energy-Poland; AECL-Canada; Hungarian Academy of Sciences KFKI Atomic Energy Research Inst.; Japan Atomic Energy Agency; Nuclear Power Inst. of China; Kyoto Univ. Research Reactor Inst.

2010-03-01T23:59:59.000Z

129

AGC-2 Graphite Preirradiation Data Package  

SciTech Connect

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.

David Swank; Joseph Lord; David Rohrbaugh; William Windes

2012-10-01T23:59:59.000Z

130

Achievements: Nuclear Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Argonne National Laboratory Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

131

Reactors: Modern-Day Alchemy - Argonne's Nuclear Science and Technology  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Legacy > Reactors: Modern-Day Alchemy About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

132

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Workers Clear Reactor Shields from Brookhaven Lab Workers Clear Reactor Shields from Brookhaven Lab Recovery Act Workers Clear Reactor Shields from Brookhaven Lab American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once used to absorb neutrons produced for research. The Brookhaven National Laboratory is using $39 million from the Recovery Act to decommission the Brookhaven Graphite Research Reactor, the world's first reactor built solely for peaceful research purposes. Recovery Act Workers Clear Reactor Shields from Brookhaven Lab More Documents & Publications Brookhaven Graphite Research Reactor Workshop 2011 ARRA Newsletters Idaho Crews Overcome Challenges to Safely Dispose 1-Million-Pound Hot Cell

133

Development of Technical Nuclear Forensics for Spent Research Reactor Fuel  

E-Print Network (OSTI)

Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis method was employed to reconstruct reactor parameters from a spent fuel sample using results from a radiochemical analysis. In the inverse analysis, a reactor physics code is used as a forward model. Verification and validation of different reactor physics codes was performed for usage in the inverse analysis. The verification and validation process consisted of two parts. The first is a variance analysis of Monte Carlo reactor physics burnup simulation results. The codes used in this work are MONTEBURNS and MCNPX/CINDER. Both utilize Monte Carlo transport calculations for reaction rate and flux results. Neither code has a variance analysis that will propagate through depletion steps, so a method to quantify and understand the variance propagation through these depletion calculations was developed. The second verification and validation process consisted of comparing reactor physics code output isotopic compositions to radiochemical analysis results. A sample from an Oak Ridge Research Reactor spent fuel assembly was acquired through a drilling process. This sample was then dissolved in nitric acid and diluted in three different quantities, creating three separate samples. A radiochemical analysis was completed and the results were compared to simulation outputs at different levels ofdetail. After establishing a forward model, an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict burnup, initial 235U, and 236U enrichments. The analytic calculation results may also be used stand alone or in a database search algorithm. In this work, a reactor physics code is used as a for- ward model with the analytic results as initial conditions in a numerical optimization algorithm. In the numerical analysis, the burnup and initial uranium isotopic com- positions are reconstructed until the iterative spent fuel characteristics converge with the measured data. Upon convergence of the samples burnup and initial uranium isotopic composition, the cooling time can be reconstructed. To reconstruct cooling time, the standard decay equation is inverted and solved for time. Two methods were developed. One method uses the converged burnup and initial uranium isotopic compositions along in a reactor depletion simulation. The second method uses an isotopic signature that does not decay out of its mass bin and has a simple production chain. An example would be 137Cs which decays into the stable 137Ba. Similar results are achieved with both methods, but extended shutdown time or time away from power results in over prediction of the cooling time. The over prediction of cooling time and comparison of different burnup reconstruction isotope results are indicator signatures of extended shutdown or time away from power. Due to dynamic operation in time and function, detailed power history reconstruction for research reactors is very challenging. Frequent variations in power, repeated variable shutdown time length, and experimentation history affect the spectrum an individual assembly is burned with such that full reactor parameter reconstruction is difficult. The results from this technical nuclear forensic analysis may be used with law enforcement, intelligence data, macroscopic and microscopic sample characteristics in a process called attribution to suggest or exclude possible sources of origin for a sample.

Sternat, Matthew 1982-

2012-12-01T23:59:59.000Z

134

HAPO GRAPHITE IRRADIATION CAPSULES  

SciTech Connect

A summary is presented of the broad field of graphite irradiation capsules. The various capsule designs are considered; they include temperature- controlled and temperature-monitored capsules. The components and materials of the capsules are described. Finally, methods are given for carrying out heat trandsfer calculations in capsule design and neutron spectra calculations for correlation of radiation data from different reactors. (D.L.C.)

Helm, J.W.

1963-09-18T23:59:59.000Z

135

Reactor core design and modeling of the MIT research reactor for conversion to LEU  

SciTech Connect

Feasibility design studies for conversion of the MIT Research Reactor (MITR) to LEU are described. Because the reactor fuel has a rhombic cross section, a special input processor was created in order to model the reactor in great detail with the REBUS-PC diffusion theory code, in 3D (triangular-z) geometry. Comparisons are made of fuel assembly power distributions and control blade worth vs. axial position, between REBUS-PC results and Monte Carlo predictions from the MCNP code. Results for the original HEU core at zero burnup are also compared with measurement. These two analysis methods showed remarkable agreement. Ongoing fuel cycle studies are summarized. A status report will be given as to results thus far that affect key design decisions. Future work plans and schedules to achieve completion of the conversion are presented. (author)

Newton, Thomas H. Jr. [Nuclear Reactor Laboratory, Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Olson, Arne P.; Stillman, John A. [RERTR Program, Argonne National Laboratory, Argonne, IL 60439 (United States)

2008-07-15T23:59:59.000Z

136

Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors  

SciTech Connect

The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors.Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat.The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

Radulescu, Laura ['Horia Hulubei' National Institute of Nuclear Physics and Engineering, PO BOX MG-6, Bucharest 077125 (Romania); Pavelescu, Margarit [Academy of Romanian Scientists, Bucharest (Romania)

2010-01-21T23:59:59.000Z

137

Two U.S. University Research Reactors to be Converted From Highly...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

are here Home Two U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium Two U.S. University Research Reactors to be Converted...

138

Design and optimization of a high thermal flux research reactor via Kriging-based algorithm  

E-Print Network (OSTI)

In response to increasing demands for the services of research reactors, a 5 MW LEU-fueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic ...

Kempf, Stephanie Anne

2011-01-01T23:59:59.000Z

139

Friction pressure drop measurements and flow distribution analysis for LEU conversion study of MIT Research Reactor  

E-Print Network (OSTI)

The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched ...

Wong, Susanna Yuen-Ting

2008-01-01T23:59:59.000Z

140

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

DOE Green Energy (OSTI)

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

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

2003-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Solid State Reactor Final Report  

DOE Green Energy (OSTI)

The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas of research were undertaken: (1) establishing the design and safety-related basis via neutronic and reactor control assessments with the graphite foam as heat transfer medium; (2) evaluating the thermal performance of the graphite foam for heat removal, reactor stability, reactor operations, and overall core thermal characteristics; (3) characterizing the physical properties of the graphite foam under normal and irradiated conditions to determine any effects on structure, dimensional stability, thermal conductivity, and thermal expansion; and (4) developing a power conversion system design to match the reactor operating parameters.

Mays, G.T.

2004-03-10T23:59:59.000Z

142

Early Exploration - Reactors designed/built by Argonne National Laboratory  

NLE Websites -- All DOE Office Websites (Extended Search)

Early Exploration Early Exploration About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

143

Behavior of Chlorine-36 and Tritium in Irradiated Graphite Wastes  

Science Conference Proceedings (OSTI)

This report examines the international data on the formation and distribution of 36Cl and 3H in graphite moderators in the context of the treatment and/or disposal of the material upon reactor decommissioning. Organizations in France have made major contributions to work in the field of 36Cl, and the review also considers work from the UK, USA and Ukraine.BackgroundThe Electric Power Research Institute (EPRI) has conducted a ...

2012-11-30T23:59:59.000Z

144

Graphite in Science and Nuclear Technique  

E-Print Network (OSTI)

The monograph is devoted to the application of graphite and graphite composites in science and technology. The structure and electrical properties, the technological aspects of production of high-strength synthetic graphites, the dynamics of the graphite destruction, traditionally used in the nuclear industry are discussed. It is focuses on the characteristics of graphitization and properties of graphite composites based on carbon isotope 13C. The book is based, generally, on the original results, and concentrated on the actual problems of application and testing of graphite materials in modern nuclear physics, in scientific and technical applications. For scientists and engineers specializing in nuclear physics and engineering, physics of nuclear reactors, condensed matter, for undergraduate, graduate and post-graduate students of universities physical specialties.

Zhmurikov, E I; Pokrovsky, A S; Harkov, D V; Dremov, V V; Samarin, S I

2013-01-01T23:59:59.000Z

145

Graphite in Science and Nuclear Technique  

E-Print Network (OSTI)

The monograph is devoted to the application of graphite and graphite composites in science and technology. The structure and electrical properties, the technological aspects of production of high-strength synthetic graphites, the dynamics of the graphite destruction, traditionally used in the nuclear industry are discussed. It is focuses on the characteristics of graphitization and properties of graphite composites based on carbon isotope 13C. The book is based, generally, on the original results, and concentrated on the actual problems of application and testing of graphite materials in modern nuclear physics, in scientific and technical applications. For scientists and engineers specializing in nuclear physics and engineering, physics of nuclear reactors, condensed matter, for undergraduate, graduate and post-graduate students of universities physical specialties.

E. I. Zhmurikov; I. A. Bubnenkov; A. S. Pokrovsky; D. V. Harkov; V. V. Dremov; S. I. Samarin

2013-07-07T23:59:59.000Z

146

SUPPLEMENT ANALYSIS OF FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL TRANSPORTATION ALONG OTHER THAN~. PRESENTATIVE ROUTE FROM CONCORD NAVAL WEAPO~~ STATION TO IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LADORA TORY Introduction The Department of Energy is planning to transport foreign research reactor spent nuclear fuel by rail from the Concord Naval Weapons Station (CNWS), Concord, California, to the Idaho National Engineering and Environmental Laboratory (INEEL). The environmental analysis supporting the decision to transport, by rail or truck, foreign research reactor spent nuclear fuel from CNWS to the INEEL is contained in +he Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliftration Policy Concerning Foreign Research Reactor

147

Advanced sodium fast reactor accident source terms : research needs.  

Science Conference Proceedings (OSTI)

An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France] IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH] Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan] Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France] Institute for Energy Petten, Saint-Paul-lez-Durance, France

2010-09-01T23:59:59.000Z

148

Advanced sodium fast reactor accident source terms : research needs.  

SciTech Connect

An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France] IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH] Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan] Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France] Institute for Energy Petten, Saint-Paul-lez-Durance, France

2010-09-01T23:59:59.000Z

149

Electrical Resistance of Graphitic and Graphitized Cathode ...  

Science Conference Proceedings (OSTI)

The electrical resistance of graphitic and graphitized cathode materials before and after electrolysis was also measured at temperatures from 30C to 965C. An ...

150

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

UPTON, N.Y. - American Recovery and Reinvestment Act UPTON, N.Y. - American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once used to absorb neutrons produced for research. The Brookhaven National Laboratory is using $39 million from the Recovery Act to decommission the Brookhaven Graphite Research Reactor, the world's first reactor built solely for peaceful research purposes. The decommissioning is slated for completion later this year and will end Office of Environmental Management legacy cleanup activities at the Lab. The neutron shields were located on the north and south sides of a 700-ton graphite pile. The three-inch-thick shields absorbed neutrons that escaped from the graphite pile. The shields also limited movement of the pile when the reactor was in opera-

151

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

Science Conference Proceedings (OSTI)

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

Snoj, L. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Sklenka, L.; Rataj, J. [Dept. of Nuclear Reactor, Czech Technical Univ. in Prague, V Holesovickach 2, 180 00 Prague 8 (Czech Republic); Boeck, H. [Vienna Univ. of Technology/Atominstitut, Stadionallee 2, 1020 Vienna (Austria)

2012-07-01T23:59:59.000Z

152

Decommissioning of the Austrian 10 MW Research Reactor, Results and Lessons learned Paper  

SciTech Connect

After the decision to shut down the 10 MW ASTRA-MTR Research Reactor was reached in May 1998, the possible options and required phases for decommissioning and removal of the radioactive components were evaluated in a decommissioning study. To support the decisions at each phase, an estimate of the activity inventory in the various parts of the reactor and the waste volume to be expected was performed. Of the possible options an immediate dismantling to phase 1 of IAEA Technical Guide Lines after the immediately following, continued dismantling to phase 2 of these guide lines was identified as the most reasonable and under the auspices optimum choice. The actual decommissioning work on the ASTRA-Reactor began in January 2000 after its final shutdown on July 31, 1999. Preliminary evaluations of the activity inventory gave an estimated amount of 320 kg of intermediate level waste, of about 60 metric tons of contaminated and another 100 metric tons of activated low level radioactive waste. The activities were roughly estimated to be at 200 TBq in the intermediate level and 6 GBq in the low level. The structure of the decommissioning process was decided against cost-, time- and risk-optimization following the basic layout of the main tasks, e.g. the removing of the fuel, the recovering and the treatment of the intermediate level activities in the vicinity of the core, the handling and conditioning of the neutron exposed graphite and the Beryllium-elements. As an example, the dismantling of approx. 1400 metric tons of the biological shield is described in more detail from the determination of the dismantling technique to the clearing procedures and the deposition. The process of dismantling of the biological shield is presented in fast motion. The dismantling of the pump-room installations of the primary loop, the processing of the contaminated or activated metals, the dismantling of the ventilation system and the radiological clearance of the reactor building was done under optimized conditions and is explained in the following. Spent fuel was generally delivered to the US Department of Energy - DOE in several shipments over the operational time of the ASTRA reactor. With the last shipment in May 2001 all the remaining spent fuel elements out of the ASTRA reactor consignment were transferred to DOE. To reduce waste from concrete shielding, German regulations Dt.StrSchV, annex IV, table 1, two clearance values referring to 'clearance restricted for permanent deposit' and to a clearance for unrestricted re-use were used. In order to reduce the amount of an estimated 60 tons of slightly contaminated metals, it was determined that introducing re-melting procedures were the most economical way. To obtain radiological clearance of the reactor building, compliance with the release limits according to Austrian Radiation Protection Ordinance had to be proved to the regulatory body. There, in general, the limits for unrestricted release were defined as a maximum dose rate of 10 {mu}Sv effective for an individual person per year. The results of the regular yearly medical examinations of the staff indicated no influence of the work related to decommissioning. The readings of the personal dosimeters over the entire project amounted to a total of 85.6 mSv, averaging to 1.07 mSv per year and person. After finishing the decommissioning process, the material balance showed 89.6 % for unrestricted reuse, 6.6 % for conventional mass-dumping and 3.8 % of ILW and LLW. The project was covered by an extensive documentation. All operations within NES followed ISO 9000 quality insurance standards. Experiences and knowledge were presented to and shared with the community, e.g. AFR and IAEA throughout the project. (authors)

Hillebrand, G.; Meyer, F. [Nuclear Engineering Seibersdorf GmbH (NES), Seibersdorf, Austria, Europe (Austria)

2008-07-01T23:59:59.000Z

153

Manhattan Project: Production Reactor (Pile) Design, Met Lab, 1942  

Office of Scientific and Technical Information (OSTI)

Schematic of the X-10 Graphite Reactor, Oak Ridge PRODUCTION REACTOR (PILE) DESIGN Schematic of the X-10 Graphite Reactor, Oak Ridge PRODUCTION REACTOR (PILE) DESIGN (Met Lab, 1942) Events > The Plutonium Path to the Bomb, 1942-1944 Production Reactor (Pile) Design, 1942 DuPont and Hanford, 1942 CP-1 Goes Critical, December 2, 1942 Seaborg and Plutonium Chemistry, 1942-1944 Final Reactor Design and X-10, 1942-1943 Hanford Becomes Operational, 1943-1944 By 1942, scientists had established that some of the uranium exposed to radioactivity in a reactor (pile) would eventually decay into plutonium, which could then be separated by chemical means from the uranium. Important theoretical research on this was ongoing, but the work was scattered at various universities from coast to coast. In early 1942, Arthur Compton arranged for all pile research to be moved to the Met Lab at the University of Chicago.

154

Role of research reactors in training of NPP personnel with special focus on training reactor VR-1  

Science Conference Proceedings (OSTI)

Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training program are demonstrated. (authors)

Sklenka, L.; Rataj, J.; Frybort, J.; Huml, O. [Dept. of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical Univ. in Prague, V Holesovickach 2, Prague 8, 180 00 (Czech Republic)

2012-07-01T23:59:59.000Z

155

Modular Pebble Bed Reactor Project, University Research Consortium Annual Report  

Science Conference Proceedings (OSTI)

This project is developing a fundamental conceptual design for a gas-cooled, modular, pebble bed reactor. Key technology areas associated with this design are being investigated which intend to address issues concerning fuel performance, safety, core neutronics and proliferation resistance, economics and waste disposal. Research has been initiated in the following areas: Improved fuel particle performance Reactor physics Economics Proliferation resistance Power conversion system modeling Safety analysis Regulatory and licensing strategy Recent accomplishments include: Developed four conceptual models for fuel particle failures that are currently being evaluated by a series of ABAQUS analyses. Analytical fits to the results are being performed over a range of important parameters using statistical/factorial tools. The fits will be used in a Monte Carlo fuel performance code, which is under development. A fracture mechanics approach has been used to develop a failure probability model for the fuel particle, which has resulted in significant improvement over earlier models. Investigation of fuel particle physio-chemical behavior has been initiated which includes the development of a fission gas release model, particle temperature distributions, internal particle pressure, migration of fission products, and chemical attack of fuel particle layers. A balance of plant, steady-state thermal hydraulics model has been developed to represent all major components of a MPBR. Component models are being refined to accurately reflect transient performance. A comparison between air and helium for use in the energy-conversion cycle of the MPBR has been completed and formed the basis of a masters degree thesis. Safety issues associated with air ingress are being evaluated. Post shutdown, reactor heat removal characteristics are being evaluated by the Heating-7 code. PEBBED, a fast deterministic neutronic code package suitable for numerous repetitive calculations has been developed. Use of the code has focused on scoping studies for MPBR design features and proliferation issues. Publication of an archival journal article covering this work is being prepared. Detailed gas reactor physics calculations have also been performed with the MCNP and VSOP codes. Furthermore, studies on the proliferation resistance of the MPBR fuel cycle has been initiated using these code Issues identified during the MPBR research has resulted in a NERI proposal dealing with turbo-machinery design being approved for funding beginning in FY01. Two other NERI proposals, dealing with the development of a burnup meter and modularization techniques, were also funded in which the MIT team will be a participant. A South African MPBR fuel testing proposal is pending ($7.0M over nine years).

Petti, David Andrew

2000-07-01T23:59:59.000Z

156

SLAC National Accelerator Laboratory - LCLS Graphite Experiment...  

NLE Websites -- All DOE Office Websites (Extended Search)

LCLS Graphite Experiment Poses New Questions for Researchers By Glenn Roberts Jr. May 21, 2012 In experiments at SLAC National Accelerator Laboratory, a powerful X-ray laser...

157

EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

2: Urgent-Relief Acceptance of Foreign Research Reactor Spent 2: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel SUMMARY This EA evaluates the environmental impacts of a proposal to accept 409 spent fuel elements from eight foreign research reactors in seven European countries. The spent fuel would be shipped across the ocean in spent fuel transportation casks from the country of origin to one or more United States eastern seaboard ports. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD April 22, 1994 EA-0912: Finding of No Significant Impact Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel April 22, 1994 EA-0912: Final Environmental Assessment Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

158

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)

Post, R.G.

1963-05-01T23:59:59.000Z

159

FUEL ELEMENTS FOR THE ARGONNE ADVANCED RESEARCH REACTOR  

SciTech Connect

The core design and the fuel element concept for the high-flux Argonne Advanced Research Reactor are presented. The core is cooled and moderated by light water and utilizes beryllium as a reflector. The fuel element assembly is rhomboidal in cross section and consists of 27 plates fastened together at their edges by dovetailed locking keys, and at each end by end fittings. Each fuel plate is 40 mils thick and contains a uniform dispersion of highly enriched UO/ sub 2/ particles, up to a maximum of 37 wt%, in a matrix of sintered stainless steel powder. A 5 mil thick stainless steel cladding is metallurgically bonded to each side of the fueled matrix. (N.W.R.)

Adolph, N.R.; Silberstein, M.S.; Weinstein, A.

1962-01-01T23:59:59.000Z

160

Foreign Research Reactor Spent Nuclear Fuel Acceptance Program  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Global Threat Reduction Initiative: Global Threat Reduction Initiative: U.S. Nuclear Remove Program Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance 2007 DOE TEC Meeting Chuck Messick DOE/NNSA/SRS 2 Contents * Program Objective and Policy * Program implementation status * Shipment Information * Operational Logistics * Lessons Learned * Conclusion 3 U.S. Nuclear Remove Program Objective * To play a key role in the Global Threat Reduction Remove Program supporting permanent threat reduction by accepting program eligible material. * Works in conjunction with the Global Threat Reduction Convert Program to accept program eligible material as an incentive to core conversion providing a disposition path for HEU and LEU during the life of the Acceptance Program. 4 Reasons for the Policy

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

MCNP-model for the OAEP Thai Research Reactor  

SciTech Connect

An MCNP input was prepared for the Thai Research Reactor, making extensive use of the MCNP geometry`s lattice feature that allows a flexible and easy rearrangement of the core components and the adjustment of the control elements. The geometry was checked for overdefined or undefined zones by two-dimensional plots of cuts through the core configuration with the MCNP geometry plotting capabilities, and by a three-dimensional view of the core configuration with the SABRINA code. Cross sections were defined for a hypothetical core of 67 standard fuel elements and 38 low-enriched uranium fuel elements--all filled with fresh fuel. Three test calculations were performed with the MCNP4B-code to obtain the multiplication factor for the cases with control elements fully inserted, fully withdrawn, and at a working position.

Gallmeier, F.X.; Tang, J.S.; Primm, R.T. III

1998-06-01T23:59:59.000Z

162

Sodium fast reactor fuels and materials : research needs.  

SciTech Connect

An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

Denman, Matthew R.; Porter, Douglas (Idaho National Laboratory, Idaho Falls, ID); Wright, Art (Argonne National Laboratory Argonne, IL); Lambert, John (Argonne National Laboratory Argonne, IL); Hayes, Steven (Idaho National Laboratory, Idaho Falls, ID); Natesan, Ken (Argonne National Laboratory Argonne, IL); Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Garner, Frank (Radiation Effects Consulting. Richland, WA); Walters, Leon (Advanced Reactor Concepts, Idaho Falls, ID); Yacout, Abdellatif (Argonne National Laboratory Argonne, IL)

2011-09-01T23:59:59.000Z

163

High uranium density dispersion fuel for the reduced enrichment of research and test reactors program.  

E-Print Network (OSTI)

??This work describes the fabrication of a high uranium density fuel for the Reduced Enrichment of Research and Test Reactors Program. In an effort to (more)

[No author

2006-01-01T23:59:59.000Z

164

Final Site-Specific Decommissioning Inspection Report for the University of Washington Research and Test Reactor  

SciTech Connect

Report of site-specific decommissioning in-process inspection activities at the University of Washington Research and Test Reactor Facility.

Sarah Roberts

2006-10-18T23:59:59.000Z

165

The inverse kinetics method and PID compensation of the Annular Core Research Reactor.  

E-Print Network (OSTI)

??This thesis explores the development of a model describing the Annular Core Research Reactor (ACRR), the application of the inverse kinetics method to calculate the (more)

Garnas, Benjamin

2008-01-01T23:59:59.000Z

166

Epithermal beam development at the BMRR (Brookhaven Medical Research Reactor): Dosimetric evaluation  

SciTech Connect

The utilization of an epithermal neutron beam for neutron capture therapy (NCT) is desirable because of the increased tissue penetration relative to a thermal neutron beam. Over the past few years, modifications have been and continue to be made at the Brookhaven Medical Research Reactor (BMRR) by changing its filter components to produce an optimal epithermal beam. An optimal epithermal beam should contain a low fast neutron contamination and no thermal neutrons in the incident beam. Recently a new moderator for the epithermal beam has been installed at the epithermal port of the BMRR and has accomplished this task. This new moderator is a combination of alumina (Al{sub 2}O{sub 3}) bricks and aluminum (Al) plates. A 0.51 mm thick cadmium (Cd) sheet has reduced the thermal neutron intensity drastically. Furthermore, an 11.5 cm thick bismuth (Bi) plate installed at the port surface has reduced the gamma dose component to negligible levels. Foil activation techniques have been employed by using bare gold and cadmium-covered gold foil to determine thermal as well as epithermal neutron fluence. Fast neutron fluence has been determined by indium foil counting. Fast neutron and gamma dose in soft tissue, free in air, is being determined by the paired ionization chamber technique, using tissue equivalent (TE) and graphite chambers. Thermoluminescent dosimeters (TLD-700) have also been used to determine the gamma dose independently. This paper describes the methods involved in the measurements of the above mentioned parameters. Formulations have been developed and the various corrections involved have been detailed. 12 refs.

Saraf, S.K.; Fairchild, R.G.; Kalef-Ezra, J.; Laster, B.H.; Fiarman, S.; Ramsey, E. (Brookhaven National Lab., Upton, NY (USA); Ioannina Univ. (Greece); Brookhaven National Lab., Upton, NY (USA); State Univ. of New York, Stony Brook, NY (USA). Health Science Center)

1989-08-24T23:59:59.000Z

167

Sodium fast reactor safety and licensing research plan. Volume I.  

SciTech Connect

This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

Sofu, Tanju (Argonne National Laboratory, Argonne, IL); LaChance, Jeffrey L.; Bari, R. (Brokhaven National Laboratory Upton, NY); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.; Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN)

2012-05-01T23:59:59.000Z

168

The initial results of research on two-step cascades in the Dalat research reactor  

E-Print Network (OSTI)

By the financial support of Vietnam Atomic Energy Commission and kind cooperation of Frank Laboratory, in the year of 2005 a measure system based on summation of amplitude pulses was established on the tangential channel of Dalat Research Reactor. After a serial of testing, the measure system was explored. In this, we would like to show the initial results were gotten with 36-Cl isotope.

Nguyen Xuan Hai; Pham Dinh Khang; Vuong Huu Tan; Ho Huu Thang; A. M. Sukhovoj; V. A. Khitrov

2013-07-11T23:59:59.000Z

169

AGC-3 Graphite Preirradiation Data Analysis Report  

SciTech Connect

This report describes the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the third Advanced Graphite Capsule (AGC-3) irradiation capsule. The AGC-3 capsule is third in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. The general design of AGC-3 test capsule is similar to the AGC-2 test capsule, material property tests were conducted on graphite specimens prior to loading into the AGC-3 irradiation assembly. However the 6 major nuclear graphite grades in AGC-2 were modified; two previous graphite grades (IG-430 and H-451) were eliminated and one was added (Mersens 2114 was added). Specimen testing from three graphite grades (PCEA, 2114, and NBG-17) was conducted at Idaho National Laboratory (INL) and specimen testing for two grades (IG-110 and NBG-18) were conducted at Oak Ridge National Laboratory (ORNL) from May 2011 to July 2013. This report also details the specimen loading methodology for the graphite specimens inside the AGC-3 irradiation capsule. The AGC-3 capsule design requires "matched pair" creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-3 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce "matched pairs" of graphite samples above and below the AGC-3 capsule elevation mid-point to provide specimens with similar neutron dose levels.

William Windes; David Swank; David Rohrbaugh; Joseph Lord

2013-09-01T23:59:59.000Z

170

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

Wigner, E.P.

1960-11-22T23:59:59.000Z

171

The Integral Fast Reactor (IFR) - Reactors designed/built by Argonne  

NLE Websites -- All DOE Office Websites (Extended Search)

Integral Fast Reactor Integral Fast Reactor About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

172

Independent Verification of Research Reactor Operation (Analysis of the Georgian IRT-M Reactor by the Isotope Ratio Method)  

SciTech Connect

The U.S. Department of Energys Office of Nonproliferation and International Security (NA-24) develops technologies to aid in implementing international nuclear safeguards. The Isotope Ratio Method (IRM) was successfully developed in 2005 2007 by Pacific Northwest National Laboratory (PNNL) and the Republic of Georgias Andronikashvili Institute of Physics as a generic technology to verify the declared operation of water-moderated research reactors, independent of spent fuel inventory. IRM estimates the energy produced over the operating lifetime of a fission reactor by measuring the ratios of the isotopes of trace impurity elements in non-fuel reactor components.The Isotope Ratio Method is a technique for estimating the energy produced over the operating lifetime of a fission reactor by measuring the ratios of the isotopes of impurity elements in non-fuel reactor components.

Cliff, John B.; Frank, Douglas P.; Gerlach, David C.; Gesh, Christopher J.; Little, Winston W.; Reid, Bruce D.; Tsiklauri, Georgi V.; Abramidze, Sh; Rostomashvili, Z.; Kiknadze, G.; Dzhavakhishvily, O.; Nabakhtiani, G.

2010-08-11T23:59:59.000Z

173

Irradiation Induced Dimensional Changes in Bulk Graphite; The theory  

E-Print Network (OSTI)

Basing on experimental data on irradiation-induced deformation of graphite we introduced a concept of diffuse domain structure developed in reactor graphite produced by extrusion. Such domains are considered as random continuous deviations of local graphite texture from the global one. We elucidate the origin of domain structure and estimate the size and the degree of orientational ordering of its domains. Using this concept we explain the well known radiation-induced size effect observed in reactor graphite. We also propose a method for converting the experimental data on shape-change of finite-size samples to bulk graphite. This method gives a more accurate evaluation of corresponding data used in estimations of reactor graphite components lifetime under irradiation.

Panyukov, S V; Arjakov, M V

2012-01-01T23:59:59.000Z

174

Irradiation Induced Dimensional Changes in Bulk Graphite; The theory  

E-Print Network (OSTI)

Basing on experimental data on irradiation-induced deformation of graphite we introduced a concept of diffuse domain structure developed in reactor graphite produced by extrusion. Such domains are considered as random continuous deviations of local graphite texture from the global one. We elucidate the origin of domain structure and estimate the size and the degree of orientational ordering of its domains. Using this concept we explain the well known radiation-induced size effect observed in reactor graphite. We also propose a method for converting the experimental data on shape-change of finite-size samples to bulk graphite. This method gives a more accurate evaluation of corresponding data used in estimations of reactor graphite components lifetime under irradiation.

S. V. Panyukov; A. V. Subbotin; M. V. Arjakov

2012-10-14T23:59:59.000Z

175

DISPERSIONS OF URANIUM CARBIDES IN ALUMINUM PLATE-TYPE RESEARCH REACTOR FUEL ELEMENTS  

DOE Green Energy (OSTI)

The technical feasibility of employing uranium carbide aluminun dispersions in aluminum-base research reactor fuel elements was investigated This study was motivated by the need to obtain higher uranium loadings in these fuel elements. Although toe MTR-type unit, containing a 13 18 wt% U-Al alloy is a proven reactor component, fabrication problems of considerable magnitude arise when attempts are made to increase the uranium investment in the alloy to more than 25 wt.%. Au approach to these fabrication difficulties is to select a compound with significantly higher density tban UAl/sub 4/ or UAl/sub 3/ compounds of the alloy system which when dispersed in aluminum powder, will reduce the volume occupied by the brittle, fissile phase. The uranium carbides, with densities ranging from 11.68 to 13.63 g/cm/sup 3/), appear to be suited for this application and were selected for development as a fuel material for aluminum-base dispersions. Studies were conducted at 580 to 620 deg C to determine the chemical compatibility of carbides with aluminum in sub-size cold- pressed comparts as well as in full-size fabricated fuel plates. Procedures were also developed to prepare uranium carbides, homogernously disperse the compounds in aluminum, roll clad the dispersions to form composite plates, and braze the plates into fuel assemblies. Corrosion tests of the fuel material were conducted in 20 and 60 deg C water to determine the integrity of the fuel material in the event of sin inadventent cladding failure. In addition, specimens were prepared to evaluate penformance under extensive irradiation Prior to studying the uranium carbide-aluminum system, methods for preparing the carbides were investigated. Are melting uranium and carnon was satisfactory for obtaining small quantities of various carbides. Later, reaction of graphite with UO/sub 2/ was successfully employed in the preparation of large quantities of UC/sub 2/, Studies of the chemical compatibility of cold-pressed compacts containing 50 wt% uranium carbide dispersed in aluminum revealed a marked trend toward stebifity as the carbon content of the uranium carbide increased from 446 to 9.20% C. Severe volume increases occurred in monocarbide dispersions with attendant formation of large quantities of the uranium-allumnim inter-metallic compounds. Dicarbide dispersions, on the other band, exhibited negligible reaction with aluminum after extended periods at 580 and 620 deg C. However, it was demonstrated that hydrogen can promote a reaction in UC/sub 2/-Al compacts. The hydrogen appears to reduce the UC/sub 2/ to UC which can subsequently react with aluminum producing the previously noted deleterious effects. A growth study at 605 deg C of composite fuel plates containing 59 wt.% UC/sub 2/ revealed insignificant changes within processing periods envisioned for fuel element processing. However, plate elongations as high as 2.5% were observed after 100 hr at this temperature. Severe blistering which occurred on fuel plates fabricated in the initial stages of the investigation was attributed to gaseous hydrocarbons, and the condition was ellminated by vacuum degasification of cold-pressed compacts. With the exception of the degasification requirement, procedures for manufacturing UC- bearing fuel elements were identical to those specified for the Geneva Conference Reactor fuel elements. Dispersions of uranium dicarbide corroded catastrophically in 20 and 60 deg C water, thus limiting the application of this material However, spocimens were prepared and insented in the MTR to evaluate the irradiation behavior of this fuel because of its potential application in onganic- cooled reactors. (auth)

Thurber, W.C.; Beaver, R.J.

1959-11-19T23:59:59.000Z

176

Development and transfer of fuel fabrication and utilization technology for research reactors  

SciTech Connect

Approximately 300 research reactors supplied with US-enriched uranium are currently in operation in about 40 countries, with a variety of types, sizes, experiment capabilities and applications. Despite the usefulness and popularity of research reactors, relatively few innovations in their core design have been made in the last fifteen years. The main reason can be better understood by reviewing briefly the history of research reactor fuel technology and enrichment levels. Stringent requirements on the enrichment of the uranium to be used in research reactors were considered and a program was launched to assist research reactors in continuing their operation with the new requirements and with minimum penalties. The goal of the new program, the Reduced Enrichment Research and Test Reactor (RERTR) Program, is to develop the technical means to utilize LEU instead of HEU in research reactors without significant penalties in experiment performance, operating costs, reactor modifications, and safety characteristics. This paper reviews briefly the RERTR Program activities with special emphasis on the technology transfer aspects of interest to this conference.

Travelli, A.; Domagala, R.F.; Matos, J.E.; Snelgrove, J.L.

1982-01-01T23:59:59.000Z

177

U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization  

SciTech Connect

The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies.

Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

2004-10-03T23:59:59.000Z

178

Nuclear reactor and materials science research: Technical report, May 1, 1985-September 30, 1986  

Science Conference Proceedings (OSTI)

Throughout the 17-month period of its grant, May 1, 1985-September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The reactor was operated 4115 hours during FY 1986 and for 6080 hours during the entire 17-month period, an average of 82 hours per week. Utilization of the reactor during that period may be classified as follows: neutron beam tube research; nuclear materials research and development; radiochemistry and trace analysis; nuclear medicine; radiation health physics; computer control of reactors; dose reduction in nuclear power reactors; reactor irradiations and services for groups outside MIT; MIT Research Reactor. Data on the above utilization for FY 1986 show that the MIT Nuclear Reactor Laboratory (NRL) engaged in joint activities with nine academic departments and interdepartmental laboratories at MIT, the Charles Stark Draper Laboratory in Cambridge, and 22 other universities and nonprofit research institutions, such as teaching hospitals.

Not Available

1987-05-11T23:59:59.000Z

179

The Optimized Integration of the Decontamination Plan and the Radwaste Management Plan into Decommissioning Plan to the VVR-S Research Reactor from Romania  

SciTech Connect

The paper presents the progress of the Decontamination Plan and Radioactive Waste Management Plan which accompanies the Decommissioning Plan for research reactor VVR-S located in Magurele, Ilfov, near Bucharest, Romania. The new variant of the Decommissioning Plan was elaborated taking into account the IAEA recommendation concerning radioactive waste management. A new feasibility study for VVR-S decommissioning was also elaborated. The preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as rehabilitation of the existing Radioactive Waste Treatment Plant and the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and reusing of cleaned reactor building is envisaged. An inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the IAEA assistance. Environmental concerns are a part of the radioactive waste management strategy. In conclusion: The current version 8 of the Draft Decommissioning Plan which include the Integrated concept of Decontamination and Decommissioning and Radwaste Management, reflects the substantial work that has been incorporated by IFIN-HH in collaboration with SITON, which has resulted in substantial improvement in document The decommissioning strategy must take into account costs for VVR-S Reactor decommissioning, as well as costs for much needed refurbishments to the radioactive waste treatment plant and the Baita-Bihor waste disposal repository. Several improvements to the Baita-Bihor repository and IFIN-HH waste treatment facility were proposed. The quantities and composition of the radioactive waste generated by VVR-S Reactor dismantling were again estimated by streams and the best demonstrated practicable processing solution was proposed. The estimated quantities of materials to be managed in the near future raise some issues that need to be solved swiftly, such as treatment of aluminum and lead and graphite management. It is envisaged that these materials to be treated to Subsidiary for Nuclear Research (SCN) Pitesti. (authors)

Barariu, G. [National Authority for Nuclear Activity-Subsidiary of Technology and Engineering for Nuclear Projects (Romania)

2008-07-01T23:59:59.000Z

180

Carbon-14 in Irradiated Graphite Waste  

Science Conference Proceedings (OSTI)

This report examines the international data on the formation and distribution of 14C in graphite moderators in the context of the treatment and/or disposal of the material upon reactor decommissioning. International organizations from the United States, France, Germany, Italy, Lithuania, and the United Kingdom collaborated in this program. This report provides an informed and improved understanding of the formation and behavior of 14C in irradiated graphite to determine where agreement or residual differ...

2010-12-20T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
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181

Monitoring and Control Research Using a University Reactor and SBWR Test-Loop  

Science Conference Proceedings (OSTI)

The existing hybrid simulation capability of the Penn State Breazeale nuclear reactor was expanded to conduct research for monitoring, operations and control. Hybrid simulation in this context refers to the use of the physical time response of the research reactor as an input signal to a real-time simulation of power-reactor thermal-hydraulics which in-turn provides a feedback signal to the reactor through positioning of an experimental changeable reactivity device. An ECRD is an aluminum tube containing an absorber material that is positioned in the central themble of the reactor kinetics were used to expand the hybrid reactor simulation (HRS) capability to include out-of-phase stability characteristics observed in operating BWRs.

Robert M. Edwards

2003-09-28T23:59:59.000Z

182

Gas release driven dynamics in research reactors piping  

SciTech Connect

Analysis of the physical and chemical processes of radiolysis gas production, air absorption, diffusion controlled gas release and transport in the coolant cleaning system of the research reactor FRM II, which is now being in routine power operation in Munich, Germany, lead to the following conclusions: 1) The steady state pressure distribution in the siphon pipe allows that the horizontal part of the siphon pipe is filled with air. The air is isolated by about 1 m water column from the main pipe of the coolant cleaning system (CCS). This is a stable steady state. It has two positive impacts on the normal operation of the CCS: (a) there is effectively no bypass flow; (b) The air can not be transported through the pipe and therefore no deterioration of the pump performance is expected from the function of the siphon pipe. 2) Radiolysis gas production for coolant, that initially does not contain dissolved air, does not lead to any problem for the system. The gases are dissolved in the coolant at 2.2 bar and are not released for pressures reduction to about 1 bar, which is the minimum pressure in the CCS. 3) Assuming hypothetically a radiolysis gas production for coolant, which initially does contain dissolved air close to its saturation, leads to gas slug formation and its transport up to the pump. This could reduce the pump head and could lead to distortion of the normal operation. Systematic measurement of the hydrogen in the primary system at 100% power indicated, that this state is not realized in the system. The observed H{sub 2} concentration was between 0.016 e-6 and 0.380 e-6 which is of no concern at all. (authors)

Kolev, Nikolay Ivanov; Roloff-Bock, Iris; Schlicht, Gerhard [Framatome ANP, P.O. Box 3220, D-91058, Erlangen (Germany)

2006-07-01T23:59:59.000Z

183

Risk management at the Oak Ridge National Laboratory research reactors  

SciTech Connect

In November of 1986, the High Flux Isotope Reactor (HFIR) was shut down by Oak Ridge National Laboratory (ORNL) due to a concern regarding embrittlement of the reactor vessel. A massive review effort was undertaken by ORNL and the Department of Energy (DOE). This review resulted in an extensive list of analyses and design modifications to be completed before restart could take place. The review also focused on the improvement of management practices including implementation of several of the Institute of Nuclear Power Operations (INPO) requirements. One of the early items identified was the need to perform a Probabilistic Risk Assessment (PRA) on the reactor. It was decided by ORNL management that this PRA would not be just an exercise to assess the ``bottom`` line in order to restart, but would be used to improve the overall safety of the reactor, especially since resources (both manpower and dollars) were severely limited. The PRA would become a basic safety tool to be used instead of a more standard deterministic approach to safety used in commercial reactor power plants. This approach was further reinforced, because the reactor was nearly 25 years old at this time, and the design standards and regulations had changed significantly since the original design, and many of the safety issues could not be addressed by compliance to codes and standards.

Flanagan, G.F.; Linn, M.A.; Proctor, L.D.; Cook, D.H.

1994-12-31T23:59:59.000Z

184

USA NRC/RSR Data Bank System and Reactor Safety Research Data Repository (RSRDR)  

SciTech Connect

The United States Nuclear Regulatory Commission (NRC), through its Division of Reactor Safety Research (RSR) of the Office of Nuclear Regulatory Research, has established the NRC/RSR Data Bank Program to collect, process, and make available data from the many domestic and foreign water reactor safety research programs. An increasing number of requests for data and/or calculations generated by NRC Contractors led to the initiation of the program which allows timely and direct access to water reactor safety data in a manner most useful to the user. The program consists of three main elements: data sources, service organizations, and a data repository.

Maskewitz, B.F.; Bankert, S.F.

1979-01-01T23:59:59.000Z

185

Early Argonne reactor lit the way for worldwide nuclear industry -  

NLE Websites -- All DOE Office Websites (Extended Search)

Early Argonne reactor lit the way for worldwide Early Argonne reactor lit the way for worldwide nuclear industry About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

186

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Environmental Protection Division Environmental Protection Division Home Reactor Projects Celebrating DOE's Cleanup Accomplishments (PDF) Brookhaven Graphite Research Reactor(BGRR) BGRR Overview BGRR Complex Description Decommissioning Decision BGRR Complex Cleanup Actions BGRR Documents BGRR Science & Accomplishments High Flux Beam Reactor (HFBR) HFBR Overview HFBR Complex Description Decommissioning Decision HFBR Complex Cleanup Actions HFBR Documents HFBR Science & Accomplishments Groundwater Protection Group Environmental Protection Division Contact > See also: HFBR Science & Accomplishments High Flux Beam Reactor Under the U.S. Department of Energy (DOE), the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) underwent stabilization and partial decommissioning to prepare the HFBR confinement for long-term safe

187

Thermal-hydraulic aspects of flow inversion in a research reactor  

SciTech Connect

PARET, a neutronics and thermal-hydraulics computer code, has been modified to account for natural convection in a reactor core. The code was then used to analyze the flow inversion that occurs in a reactor with heat removal by forced convection in the downward direction after a pump failure. Typical results are shown for a number of parameters. Research reactors normally operating much above ten MW are predicted to experience nucleate boiling in the event of a flow inversion. Comparison with experimental results from the Belgian BR2 reactor indicated general agreement although nucleate boiling that was analytically predicted was not noted in the BR2 data.

Smith, R.S.; Woodruff, W.L.

1986-11-01T23:59:59.000Z

188

Light Water Reactor Fuel Cladding Research and Testing | ornl.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Light Water Reactor Fuel Cladding Research Light Water Reactor Fuel Cladding Research June 01, 2013 Severe Accident Test Station ORNL is the focus point for Light Water Reactor (LWR) fuel cladding research and testing. The purpose of this research is to furnish U.S. industry (EPRI, Areva, Westinghouse), and regulators (NRC) with much-needed data supporting safe and economical nuclear power generation and used fuel management. LWR fuel cladding work is tightly integrated with ORNL accident tolerant fuel development and used fuel disposition programs thereby providing a powerful capability that couples basic materials science research with the nuclear applications research and development. The ORNL LWR fuel cladding program consists of five complementary areas of research: Accident tolerant fuel and cladding material testing under design

189

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

SciTech Connect

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

Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

2012-09-01T23:59:59.000Z

190

Two U.S. University Research Reactors to be Converted From Highly Enriched  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

U.S. University Research Reactors to be Converted From Highly U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium Two U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium April 11, 2005 - 11:34am Addthis WASHINGTON, D.C. - As part of the Bush administration's aggressive effort to reduce the amount of weapons-grade nuclear material worldwide, Secretary of Energy Samuel W. Bodman announced today that the Department of Energy (DOE) has begun to convert research reactors from using highly-enriched uranium (HEU) to low-enriched uranium fuel (LEU) at the University of Florida and Texas A&M University. This effort, by DOE's National Nuclear Security Administration (NNSA) and the Office of Nuclear Energy, Science and Technology, are the latest steps

191

Effect of reduced enrichment on the fuel cycle for research reactors  

SciTech Connect

The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel.

Travelli, A.

1982-01-01T23:59:59.000Z

192

Modeling Fission Product Sorption in Graphite Structures  

SciTech Connect

The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high- temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products on each type of graphite site. The model will include multiple simultaneous adsorbing species, which will allow for competitive adsorption effects between different fission product species and O and OH (for modeling accident conditions).

Szlufarska, Izabela [University of Wisconsin, Madison, WI (United States); Morgan, Dane [University of Wisconsin, Madison, WI (United States); Allen, Todd [University of Wisconsin, Madison, WI (United States)

2013-04-08T23:59:59.000Z

193

AGC-2 Graphite Preirradiation Data Analysis Report  

SciTech Connect

This report described the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the second Advanced Graphite Capsule (AGC-2) irradiation capsule. The AGC-2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. Similar to the AGC-1 specimen pre-irradiation examination report, material property tests were conducted on specimens from 18 nuclear graphite types but on an increased number of specimens (512) prior to loading into the AGC-2 irradiation assembly. All AGC-2 specimen testing was conducted at Idaho National Laboratory (INL) from October 2009 to August 2010. This report also details the specimen loading methodology for the graphite specimens inside the AGC-2 irradiation capsule. The AGC-2 capsule design requires matched pair creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-2 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce matched pairs of graphite samples above and below the AGC-2 capsule elevation mid-point to provide specimens with similar neutron dose levels.

William Windes; W. David Swank; David Rohrbaugh; Joseph Lord

2013-08-01T23:59:59.000Z

194

Demonstration of the reactivity constraint approach on SNL's annual core research reactor  

Science Conference Proceedings (OSTI)

This paper reports on the initial demonstration of the reactivity constraint approach and its implementing algorithm, the MIT-CSDL Non-Linear Digital Controller, on the annual core research reactor (ACCR) that is operated by the Sandia National Laboratories. This demonstration constituted the first use of reactivity constraints for the closed-loop, digital control of reactor power on a facility other than the Massachusetts Institute of Technology's (MIT's) research reactor (MITR-II). Also, because the ACRR and the MITR-II are of very different design, these trials established the generic nature of the reactivity constraint approach.

Bernard, J.A.; Kwok, K.S.; Wyant, F.J.; Thome, F.V.

1989-01-01T23:59:59.000Z

195

A Brief History i-l Research Reactors  

E-Print Network (OSTI)

stainless steel sam- ples in the High Flux Isotope Reactor (HFIR) at tem- peratures of 380 to 680" with up/cm' to balance the gas pressure were used m their calculation. A comparison of the results with HFIR and the HFIR ex- perimental data is presented in section 5. Applications of the model to various fusion designs

196

Neutronic reactor  

DOE Patents (OSTI)

A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

Lewis, Warren R. (Richland, WA)

1978-05-30T23:59:59.000Z

197

Benchmarking of Graphite Reflected Critical Assemblies of UO2  

Science Conference Proceedings (OSTI)

A series of experiments were carried out in 1963 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for use in space reactor research programs. A core containing 93.2% enriched UO2 fuel rods was used in these experiments. The first part of the experimental series consisted of 253 tightly-packed fuel rods (1.27 cm triangular pitch) with graphite reflectors [1], the second part used 253 graphite-reflected fuel rods organized in a 1.506 cm triangular pitch [2], and the final part of the experimental series consisted of 253 beryllium-reflected fuel rods with a 1.506 cm triangular pitch. [3] Fission rate distribution and cadmium ratio measurements were taken for all three parts of the experimental series. Reactivity coefficient measurements were taken for various materials placed in the beryllium reflected core. The first part of this experimental series has been evaluated for inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP) [4] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbooks, [5] and is discussed below. These experiments are of interest as benchmarks because they support the validation of compact reactor designs with similar characteristics to the design parameters for a space nuclear fission surface power systems. [6

Margaret A. Marshall; John D. Bess

2011-11-01T23:59:59.000Z

198

Materials Research Needs for Near-Term Nuclear Reactors  

Science Conference Proceedings (OSTI)

Technical Paper / NSF Workshop on the Research Needs of the Next Generation Nuclear Power Technology / Material

John R. Weeks

199

Thermal hydraulic limits analysis for the MIT Research Reactor low enrichment uranium core conversion using statistical propagation of parametric uncertainties  

E-Print Network (OSTI)

The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design from 15 to 18 plates per element, a reactor ...

Chiang, Keng-Yen

2012-01-01T23:59:59.000Z

200

Decommissioning Anddismantling Of The Research Reactor Salaspils. I. Conceptual Study And The First Results.  

E-Print Network (OSTI)

In May 1995, the Latvian government decided to shut down the Research Reactor Salaspils (SRR) and to dis pense with nuclear energy in future. The reactor is out of operation since July 1998. A conceptual study for the decommissioning of SRR has been carried out by Noell-KRC-Energie- und Umweltlechnik GmbH at 1998-1999 years. The Latvian government decided in October 26 1999 to start the direct dismantling to "green field" at 2001 year. The first decommissioning and dismantling results from preparation measures in 1999 year are presented and discussed. The main efforts was devoted to collecting and conditioning of "historical" radioactive wastes from different storages outside and inside of reactor hall. All non-radioactive equipments and materials outside of reactor buildings were free-released and dismantled for reusing and conventional disposing. Weakly contaminated materials from reactor hall were collected and removed for free-release measurements. 1.0 INTRODUCTION The res...

Andris Abramenkovs Ministry; Andris Abramenkovs; Arnis Ezergailis; State Enterprise vides Projekti; Dzintars Kalnin

2000-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report  

Science Conference Proceedings (OSTI)

This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOEs Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

2002-11-01T23:59:59.000Z

202

A brief History of Neutron Scattering at the Oak Ridge High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

Neutron scattering at the Oak Ridge National Laboratory dates back to 1945 when Ernest Wollan installed a modified x-ray diffractometer on a beam port of the original graphite reactor. Subsequently, Wollan and Clifford Shull pioneered neutron diffraction and laid the foundation for an active neutron scattering effort that continued through the 1950s, using the Oak Ridge Research reactor after 1958, and, starting in 1966, the High Flux Isotope Reactor, or HFIR.

Nagler, Stephen E [ORNL; Mook Jr, Herbert A [ORNL

2008-01-01T23:59:59.000Z

203

CHARACTERIZATION OF LEAK PATHWAYS IN THE BELOW GRADE DUCTS OF THE BROOKHAVEN GRAPHITE RESEARCH REACTOR USING PERFLUOROCARBON TRACERS.  

SciTech Connect

The focus of this program was the characterization of the soils beneath the main air ducts connecting the exhaust plenums with the Fan House. The air plenums experienced water intrusion during BGRR operations and after shutdown. The water intrusions were attributed to rainwater leaks into degraded parts of the system and to internal cooling water system leaks. As part of the overall characterization efforts, a state-of-the-art gaseous perfluorocarbon tracer technology was utilized to characterize leak pathways from the ducts. This in turn suggests what soil regions under or adjacent to the ductwork should be emphasized in the characterization process. Knowledge of where gaseous tracers leak from the ducts yields a conservative picture of where water transport, out of or into, the ducts might have occurred.

HEISER,J.; SULLIVAN,T.; KALB,P.; MILIAN,L.; WILKE,R.; NEWSON,C.; LILIMPAKIS,M.

2001-04-01T23:59:59.000Z

204

MITR-III: Upgrade and relicensing studies for the MIT Research Reactor. Second annual report  

SciTech Connect

The current operating license of the MIT research reactor will expire on May 7, 1996 or possibly a few years later if the US Nuclear Regulatory Commission agrees that the license period can start with the date of initial reactor operation. Driven by the imminent expiration of the operating license, a team of nuclear engineering staff and students have begun a study of the future options for the MIT Research Reactor. These options have included the range from a major rebuilding of the reactor to its decommissioning. This document reports the results of a two year intensive activity which has been supported by a $148,000 grant from the USDOE contract Number DEFG0293ER75859, approximately $100,000 of internal MIT funds and Nuclear Engineering Department graduate student fellowships as well as assistance from international visiting scientists and engineers.

Trosman, H.G. [ed.; Lanning, D.D.; Harling, O.K.

1994-08-01T23:59:59.000Z

205

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel  

SciTech Connect

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. Results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory for LEU conversion of the GTRR are summarized. Only those parameters which could change as a result of replacing the fuel are addressed. The performance of the reactor and all safety margins with LEU fuel are expected to be about the same as those with the current HEU fuel.

Matos, J.E.; Mo, S.C.; Woodruff, W.L.

1992-01-01T23:59:59.000Z

206

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel  

Science Conference Proceedings (OSTI)

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. Results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory for LEU conversion of the GTRR are summarized. Only those parameters which could change as a result of replacing the fuel are addressed. The performance of the reactor and all safety margins with LEU fuel are expected to be about the same as those with the current HEU fuel.

Matos, J.E.; Mo, S.C.; Woodruff, W.L.

1992-12-31T23:59:59.000Z

207

Leu conversion status of U.S. research reactors: September 1996  

SciTech Connect

At the request of the Department of Energy, the RERTR Program has summarized the conversion status of research and test reactors in the United States and has made estimates of the uranium densities that would be needed to convert the reactors with power levels greater than or equal to 1 MW from Highly Enriched Uranium (HEU) (greater than or equal to 20% U-235) to Lightly Enriched Uranium (LEU) (less than 20% U-235) fuels. Detailed conversion studies for each of the reactors need to be completed in order to establish the feasibility of using LEU fuels.

Matos, J.E.

1996-09-01T23:59:59.000Z

208

Education program at the Massachusetts Institute of Technology research reactor for pre-college science teachers  

Science Conference Proceedings (OSTI)

A Pre-College Science Teacher (PCST) Seminar program has been in place at the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory for 4 yr. The purpose of the PCST program is to educate teachers in nuclear technology and to show teachers, and through them the community, the types of activities performed at research reactors. This paper describes the background, content, and results of the MIT PCST program.

Hopkins, G.R.; Fecych, W.; Harling, O.K.

1989-01-01T23:59:59.000Z

209

Using low-enriched uranium in research reactors: The RERTR program  

SciTech Connect

The goal of the RERTR program is to minimize and eventually eliminate use of highway enriched uranium (HEU) in research and test reactors. The program has been very successful, and has developed low-enriched uranium (LEU) fuel materials and designs which can be used effectively in approximately 90 percent of the research and test reactors which used HEU when the program began. This progress would not have been possible without active international cooperation among fuel developers, commercial vendors, and reactor operators. The new tasks which the RERTR program is undertaking at this time include development of new and better fuels that will allow use of LEU fuels in all research and test reactors; cooperation with Russian laboratories, which will make it possible to minimize and eventually eliminate use of HEU in research reactors throughout the world, irrespective of its origin; and development of an LEU-based process for the production of {sup 99}Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program.

Travelli, A.

1994-05-01T23:59:59.000Z

210

Why Nuclear Energy? - Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Energy: Nuclear Energy: Why Nuclear Energy? About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

211

DENSITY CONTROL IN A REACTOR  

DOE Patents (OSTI)

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

Marshall, J. Jr.

1961-10-24T23:59:59.000Z

212

Major Safety Aspects of Advanced Candu Reactor and Associated Research and Development  

Science Conference Proceedings (OSTI)

The Advanced Candu{sup R} Reactor design is built on the proven technology of existing Candu plants and on AECL's knowledge base acquired over decades of nuclear power plant design, engineering, construction and research. Two prime objectives of ACR-700TM1 are cost reduction and enhanced safety. To achieve them some new features were introduced and others were improved from the previous Candu 6 and Candu 9 designs. The ACR-700 reactor design is based on the modular concept of horizontal fuel channels surrounded by a heavy water moderator, the same as with all Candu reactors. The major novelty in the ACR-700 is the use of slightly enriched fuel and light water as coolant circulating in the fuel channels. This results in a more compact reactor design and a reduction of heavy water inventory, both contributing to a significant decrease in cost compared to Candu reactors, which employ natural uranium as fuel and heavy water as coolant. The reactor core design adopted for ACR-700 also has some features that have a bearing on inherent safety, such as negative power and coolant void reactivity coefficient. Several improvements in engineered safety have been made as well, such as enhanced separation of the safety support systems. Since the ACR-700 design is an evolutionary development of the currently operating Candu plants, limited research is required to extend the validation database for the design and the supporting safety analysis. A program of safety related research and development has been initiated to address the areas where the ACR-700 design is significantly different from the Candu designs. This paper describes the major safety aspects of the ACR-700 with a particular focus on novel features and improvements over the existing Candu reactors. It also outlines the key areas where research and development efforts are undertaken to demonstrate the effectiveness and robustness of the design. (authors)

Bonechi, M.; Wren, D.J.; Hopwood, J.M. [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

2002-07-01T23:59:59.000Z

213

Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion.  

E-Print Network (OSTI)

??The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density (more)

Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

2008-01-01T23:59:59.000Z

214

Evaluation of the thermal-hydraulic operating limits of the HEU-LEU transition cores for the MIT Research Reactor.  

E-Print Network (OSTI)

??The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment (more)

Wang, Yunzhi (Yunzhi Diana)

2009-01-01T23:59:59.000Z

215

N-Reactor Department research and development budget for FY-1967 and revision of budget for FY-1966  

SciTech Connect

This report provides the N-Reactor research and development budget for fiscal year 1967 and modifications of the budget for fiscal year 1966.

1965-04-19T23:59:59.000Z

216

Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors  

SciTech Connect

The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

1993-07-01T23:59:59.000Z

217

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor.  

SciTech Connect

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required to match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.

Bretscher, M. M.

1998-10-14T23:59:59.000Z

218

M-1 N-Reactor fuel development. Excerpts from N-Reactor Department Research and Development Budget for FY-1967 and Revision of Budget for FY-1966  

SciTech Connect

The use of metallic uranium fuel elements on a large scale in a pressurized water cooled graphite moderated power reactor environment in N-Reactor not only represents an extrapolation from previous Hanford fuel technology but it represents the only production-scale application of this concept in this country or abroad. Provision of a supporting technology is and will continue to be a major activity at Hanford as only a limited amount of directly applicable R&D will be available from other sites. The major benefit to be achieved through an extensive and continuing fuel program is attainment of the full capabilities of the N-Reactor at much reduced fuel cycle costs. The current N-Reactor fuel design represents the best engineering judgment of what is required for adequate performance. While the design is intentionally conservative, some features may not fully provide the level of performance required to sustain efficient reactor operation. Results of production-scale irradiation experience and special test irradiations will provide direction to a continuing program to correct any excesses or deficiencies in the initial fuel design. Reduction of unwarranted conservatism in the design will lower fuel fabrication costs, and correction of deficiencies will lower irradiation costs through increased time operated efficiency. Costs of this program are summarized for 1965, 66, and 67. The paper describes the scope of the program, its relationship to other programs, and technical progress in FY-1965 which included improved performance of the fuel elements even with the large number of rapid shutdowns of the reactor.

1965-04-19T23:59:59.000Z

219

Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor  

SciTech Connect

A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams.

Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon [Massachusetts Institute of Technology (United States)

2005-05-15T23:59:59.000Z

220

Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania  

SciTech Connect

In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.

K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin; C. Paunoiu; M. Ciocanescu

2010-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin  

SciTech Connect

This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy`s Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006.

Matos, J.E.

1994-12-01T23:59:59.000Z

222

Thermally Conductive Graphite Foam  

oriented graphite planes, similar to high performance carbon fibers, which have been estimated to exhibit a thermal conductivity greater than 1700 ...

223

Neutronic reactor thermal shield  

DOE Patents (OSTI)

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

Wende, Charles W. J. (West Chester, PA)

1976-06-15T23:59:59.000Z

224

MYRRHA a multi-purpose hybrid research reactor for high-tech applications  

SciTech Connect

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental accelerator driven system (ADS) in development at SCK-CEN. MYRRHA is able to work both in subcritical (ADS) as in critical mode. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for generation IV (GEN IV) systems, material developments for fusion reactors, radioisotope production and industrial applications, such as Si-doping. MYRRHA will also demonstrate the ADS full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. MYRRHA is based on the heavy liquid metal technology and so it will contribute to the development of lead fast reactor (LFR) technology and in critical mode, MYRRHA will play the role of European technology pilot plant in the roadmap for LFR. In this paper the historical evolution of MYRRHA and the rationale behind the design choices is presented and the latest configuration of the reactor core and primary system is described. (authors)

Abderrahim, H. A.; Baeten, P. [SCK CEN, Boeretang 200, 2400 Mol (Belgium)

2012-07-01T23:59:59.000Z

225

Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

Not Available

1994-04-01T23:59:59.000Z

226

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel  

SciTech Connect

This document presents information concerning: analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU; changes to technical specifications mandated by the conversion of the GTRR to low enrichment fuel; changes in the Safety Analysis Report mandated by the conversion of the GTRR to low enrichment fuel; and copies of all changed pages of the SAR and the technical specifications.

Matos, J.E.; Mo, S.C.; Woodruff, W.L.

1992-09-01T23:59:59.000Z

227

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Starr, C.

1963-01-01T23:59:59.000Z

228

Analysis of capsule HFR-B1 graphite-corrosion data  

SciTech Connect

The recently completed irradiation of capsule HFR-B1 in the high-flux reactor at the Pitten Establishment in The Netherlands provided some excellent data for fission-product release. The data were obtained under irradiation and temperature conditions close to those expected during normal operation of the Modular High-Temperature Gas-cooled Reactor (MHTGR). Some of the tests at Petten were designed to measure release of fission gases during hydrolysis of failed fuel. Hydrolysis was initiated by injecting known amounts of water vapor into the capsule sweep gas. The measured concentrations of CO and CO{sub 2} in the capsule sweep gas indicated that a non-negligible amount of graphite corrosion was also occurring during the hydrolysis tests. Hence, these measurements provide some unique data for in-pile corrosion of grade H-4541 graphite by steam. In the present report, an analysis of graphite corrosion during the Petten hydrolysis tests is described. The calculations were performed using the REACT program, which is based on an improved corrosion model. The REACT program was developed as part of a research program at the University of California, San Diego, and is in operational status in the General Atomics (GA) Production Code Library. Predictions obtained with REACT show excellent agreement with the Petten graphite-corrosion data. This good agreement indicates that the currently used correlation for the steam-graphite reaction rate, which was obtained from out-of-pile measurements, may also be used to predict in-pile corrosion with good accuracy. 8 refs., 19 figs., 2 tabs.

Richards, M.B.; Gillespie, A.G.

1991-01-01T23:59:59.000Z

229

Coating method for graphite  

DOE Patents (OSTI)

A method of limiting carbon contamination from graphite ware used in induction melting of uranium alloys is provided. The graphite surface is coated with a suspension of Y/sub 2/O/sub 3/ particles in water containing about 1.5 to 4 percent by weight sodium carboxymethylcellulose.

Banker, J.G.; Holcombe, C.E. Jr.

1975-11-06T23:59:59.000Z

230

Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors` spent fuel  

SciTech Connect

One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE`s Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE`s efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE`s activities in taking back spent fuel.

1994-03-25T23:59:59.000Z

231

Monte Carlo simulation of the Massachusetts Institute of Technology Research Reactor  

SciTech Connect

The three-dimensional continuous-energy MCNP Monte Carlo code is used to develop a versatile and accurate reactor physics model of the Massachusetts Institute of Technology Research Reactor 2 (MITR-2). The validation of the model against existing experimental data is presented. Core multiplication factors as well as fast neutron in-core flux measurements were used in the validation process. The agreement between the MCNP predictions and the experimentally determined values is very good, which indicates that the Monte Carlo model is correctly simulating the MITR-2.

Redmond, E.L. II; Yanch, J.C.; Harling, O.K. (Massachusetts Inst. of Tech., Cambridge, MA (United States). Nuclear Engineering Dept.)

1994-04-01T23:59:59.000Z

232

Neutronic safety parameters and transient analyses for potential LEU conversion of the Budapest Research Reactor.  

SciTech Connect

An initial safety study for potential LEU conversion of the Budapest Research Reactor was completed. The study compares safety parameters and example transients for reactor cores with HEU and LEU fuels. Reactivity coefficients, kinetic parameters and control rod worths were calculated for cores with HEU(36%) UAl alloy fuel and UO{sub 2}-Al dispersion fuel, and with LEU (19.75%)UO{sub 2}-Al dispersion fuel that has a uranium density of about 2.5 g/cm{sup 3}. A preliminary fuel conversion plan was developed for transition cores that would convert the BRR from HEU to LEU fuel after the process is begun.

Pond, R. B.; Hanan, N. A.; Matos, J. E.; Maraczy, C.

1999-09-27T23:59:59.000Z

233

Strategic Plan for Light Water Reactor Research and Development  

Science Conference Proceedings (OSTI)

The purpose of this strategic plan is to establish a framework that will allow the Department of Energy (DOE) and the nuclear power industry to jointly plan the nuclear energy research and development (R&D) agenda important to achieving the Nation's energy goals. This strategic plan has been developed to focus on only those R&D areas that will benefit from a coordinated government/industry effort. Specifically, this plan focuses on safely sustaining and expanding the electricity output from currently operating nuclear power plants and expanding nuclear capacity through the deployment of new plants. By focusing on R&D that addresses the needs of both current and future nuclear plants, DOE and industry will be able to take advantage of the synergism between these two technology areas, thus improving coordination, enhancing efficiency, and further leveraging public and private sector resources. By working together under the framework of this strategic plan, DOE and the nuclear industry reinforce their joint commitment to the future use of nuclear power and the National Energy Policy's goal of expanding its use in the United States. The undersigned believe that a public-private partnership approach is the most efficient and effective way to develop and transfer new technologies to the marketplace to achieve this goal. This Strategic Plan is intended to be a living document that will be updated annually.

None

2004-02-01T23:59:59.000Z

234

Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria  

SciTech Connect

The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

K. J. Allen; T. G. Apostolov; I. S. Dimitrov

2009-03-01T23:59:59.000Z

235

Research helps safeguard nuclear workers worldwide - Argonne's Historical  

NLE Websites -- All DOE Office Websites (Extended Search)

Research helps safeguard nuclear workers Research helps safeguard nuclear workers worldwide About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

236

Radiological survey support activities for the decommissioning of the Ames Laboratory Research Reactor Facility, Ames, Iowa  

SciTech Connect

At the request of the Engineering Support Division of the US Department of Energy-Chicago Operations Office and in accordance with the programmatic overview/certification responsibilities of the Department of Energy Environmental and Safety Engineering Division, the Argonne National Laboratory Radiological Survey Group conducted a series of radiological measurements and tests at the Ames Laboratory Research Reactor located in Ames, Iowa. These measurements and tests were conducted during 1980 and 1981 while the reactor building was being decontaminated and decommissioned for the purpose of returning the building to general use. The results of these evaluations are included in this report. Although the surface contamination within the reactor building could presumably be reduced to negligible levels, the potential for airborne contamination from tritiated water vapor remains. This vapor emmanates from contamination within the concrete of the building and should be monitored until such time as it is reduced to background levels. 2 references, 8 figures, 6 tables.

Wynveen, R.A.; Smith, W.H.; Sholeen, C.M.; Justus, A.L.; Flynn, K.F.

1984-09-01T23:59:59.000Z

237

A Review of Previous Research in Direct Energy Conversion Fission Reactors  

DOE Green Energy (OSTI)

From the earliest days of power reactor development, direct energy conversion was an obvious choice to produce high efficiency electric power generation. Directly capturing the energy of the fission fragments produced during nuclear fission avoids the intermediate conversion to thermal energy and the efficiency limitations of classical thermodynamics. Efficiencies of more than 80% are possible, independent of operational temperature. Direct energy conversion fission reactors would possess a number of unique characteristics that would make them very attractive for commercial power generation. These reactors would be modular in design with integral power conversion and operate at low pressures and temperatures. They would operate at high efficiency and produce power well suited for long distance transmission. They would feature large safety margins and passively safe design. Ideally suited to production by advanced manufacturing techniques, direct energy conversion fission reactors could be produced more economically than conventional reactor designs. The history of direct energy conversion can be considered as dating back to 1913 when Moseleyl demonstrated that charged particle emission could be used to buildup a voltage. Soon after the successful operation of a nuclear reactor, E.P. Wigner suggested the use of fission fragments for direct energy conversion. Over a decade after Wigner's suggestion, the first theoretical treatment of the conversion of fission fragment kinetic energy into electrical potential appeared in the literature. Over the ten years that followed, a number of researchers investigated various aspects of fission fragment direct energy conversion. Experiments were performed that validated the basic physics of the concept, but a variety of technical challenges limited the efficiencies that were achieved. Most research in direct energy conversion ceased in the US by the late 1960s. Sporadic interest in the concept appears in the literature until this day, but there have been no recent significant programs to develop the technology.

DUONG,HENRY; POLANSKY,GARY F.; SANDERS,THOMAS L.; SIEGEL,MALCOLM D.

1999-09-22T23:59:59.000Z

238

REACTOR  

DOE Patents (OSTI)

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

Roman, W.G.

1961-06-27T23:59:59.000Z

239

U.S. Department of Energy Instrumentation and Controls Technology Research for Advanced Small Modular Reactors  

Science Conference Proceedings (OSTI)

Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD&D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, key DOE programs have substantial ICHMI RD&D elements to their respective research portfolio. This article describes current ICHMI research to support the development of advanced small modular reactors.

Wood, Richard Thomas [ORNL

2012-01-01T23:59:59.000Z

240

NEW METHOD OF GRAPHITE PREPARATION  

DOE Patents (OSTI)

BS>A method is described for producing graphite objects comprising mixing coal tar pitch, carbon black, and a material selected from the class comprising raw coke, calcined coke, and graphite flour. The mixture is placed in a graphite mold, pressurized to at least 1200 psi, and baked and graphitized by heating to about 2500 deg C while maintaining such pressure. (AEC)

Stoddard, S.D.; Harper, W.T.

1961-08-29T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research  

SciTech Connect

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue Universitys Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called Users Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. Users week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

John Jackson; Todd Allen; Frances Marshall; Jim Cole

2013-03-01T23:59:59.000Z

242

A neutronic feasibility study for LEU conversion of the Budapest research reactor.  

SciTech Connect

A neutronic feasibility study for conversion of the Budapest Research Reactor (BRR) from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with the KFKI Atomic Energy Research Institute in Hungary. Comparisons were made of the reactor performance with the current HEU (36%) fuel and with a proposed LEU (19.75%) fuel. Cycle lengths, thermal neutron fluxes, and rod worths were calculated in equilibrium-type cores for each type of fuel. Relative to the HEU fuel, the LEU fuel has up to a 50% longer fuel cycle length, but a 7-10% smaller thermal neutron flux in the experiment locations. The rod worths are smaller with the LEU fuel, but are still large enough to easily satisfy the BRR shutdown margin criteria. Irradiation testing of four VVR-M2 LEU fuel assemblies that are nearly the same as the proposed BRR LEU fuel assemblies is currently in progress at the Petersburg Nuclear Physics Institute.

Pond, R. B.

1998-10-16T23:59:59.000Z

243

DEVELOPMENT OF A FAST-RELEASE ELECTRO-MAGNET FOR POOL-TYPE RESEARCH REACTORS  

SciTech Connect

Experimental data and observations of the design parameters and physical configurations which lead to faster-release cylindrical flat-faced electromagnets are given. Detailed drawings and operating characteristics are presented for an electromagnet suited to the requirements of a pool-type research reactor using either gravity drop or additional accelerating force. This electromagnet embodies the experimentally developed techniques for attaining fast release. (auth)

Michelson, C.

1958-02-14T23:59:59.000Z

244

Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report  

SciTech Connect

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

1982-03-01T23:59:59.000Z

245

A simple setup for neutron tomography at the Portuguese Nuclear Research Reactor  

E-Print Network (OSTI)

A simple setup for neutron radiography and tomography was recently installed at the Portuguese Research Reactor. The objective of this work was to determine the operational characteristics of the installed setup, namely the irradiation time to obtain the best dynamic range for individual images and the spatial resolution. The performance of the equipment was demonstrated by imaging a fragment of a 17th century decorative tile.

M. A. Stanojev Pereira; J. G. Marques; R. Pugliesi

2012-05-15T23:59:59.000Z

246

Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies  

SciTech Connect

This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

Pond, R.B.; Matos, J.E.

1996-12-31T23:59:59.000Z

247

CIVILIAN POWER REACTOR PROGRAM. PART II. ECONOMIC POTENTIAL AND DEVELOPMENT PROGRAM AS OF 1959  

SciTech Connect

The status of technology of nuclear power reactors in 1959 is reviewed. General research and engineering development activities are discussed. The reactors considered include the pressurized water, boiling water, light water moderated superheat, organic cooled, sodium graphite, gas cooled enriched fuel, gas cooled natural uranium, fast breeder, aqueous homogeneous, and heavy water. Power costs are compared with the cost of power from conventional plants. (C.H.)

1960-01-01T23:59:59.000Z

248

Reactor physics teaching and research in the Swiss nuclear engineering master  

Science Conference Proceedings (OSTI)

Since 2008, a Master of Science program in Nuclear Engineering (NE) has been running in Switzerland, thanks to the combined efforts of the country's key players in nuclear teaching and research, viz. the Swiss Federal Inst.s of Technology at Lausanne (EPFL) and at Zurich (ETHZ), the Paul Scherrer Inst. (PSI) at Villigen and the Swiss Nuclear Utilities (Swissnuclear). The present paper, while outlining the academic program as a whole, lays emphasis on the reactor physics teaching and research training accorded to the students in the framework of the developed curriculum. (authors)

Chawla, R. [Swiss Federal Inst. of Technology EPFL, CH-1015 Lausanne (Switzerland); Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland)

2012-07-01T23:59:59.000Z

249

Carbon Characterization Laboratory Readiness to Receive Irradiated Graphite Samples  

SciTech Connect

The Carbon Characterization Laboratory (CCL) is located in Labs C19 and C20 of the Idaho National Laboratory Research Center. The CCL was established under the Next Generation Nuclear Plant Project to support graphite and ceramic composite research and development activities. The research conducted in this laboratory will support the Advanced Graphite Creep experimentsa major series of material irradiation experiments within the Next Generation Nuclear Plant Graphite program. The CCL is designed to characterize and test low activated irradiated materials such as high purity graphite, carbon-carbon composites, silicon-carbide composite, and ceramic materials. The laboratory is fully capable of characterizing material properties for both irradiated and nonirradiated materials. Major infrastructural modifications were undertaken to support this new radiological facility at Idaho National Laboratory. Facility modifications are complete, equipment has been installed, radiological controls and operating procedures have been established and work management documents have been created to place the CCL in readiness to receive irradiated graphite samples.

Karen A. Moore

2011-05-01T23:59:59.000Z

250

IGORR-1: Proceedings of the first meeting of the international group on research reactors  

SciTech Connect

Many organizations, in several countries, are planning or implementing new or upgraded research reactor projects, but there has been no organized forum devoted entirely to discussion and exchange of information in this field. Over the past year or so, informal discussions resulted in widespread agreement that such a forum would serve a useful purpose. Accordingly, a proposal to form a group was submitted to the leading organizations known to be involved in projects to build or upgrade reactor facilities. Essentially all agreed to join in the formation of the International Group on Research Reactors (IGORR) and nominated a senior staff member to serve on its international organizing committee. The first IGORR meeting took place on February 28--March 2, 1990. It was very successful and well attended; some 52 scientists and engineers from 25 organizations in 10 countries participated in 2-1/2 days of open and informative presentations and discussions. Two workshop sessions offered opportunities for more detailed interaction among participants and resulted in identification of common R D needs, sources of data, and planned new facilities. Individual papers have been cataloged separately.

West, C.D. (comp.)

1990-05-01T23:59:59.000Z

251

Evaluation of the thermal-hydraulic operating limits of the HEU-LEU transition cores for the MIT Research Reactor  

E-Print Network (OSTI)

The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU) fuel. The currently selected LEU fuel design contains 18 ...

Wang, Yunzhi (Yunzhi Diana)

2009-01-01T23:59:59.000Z

252

Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion  

E-Print Network (OSTI)

The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the ...

Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

2008-01-01T23:59:59.000Z

253

NUMERICAL SIMULATION FOR MECHANICAL BEHAVIOR OF U10MO MONOLITHIC MINIPLATES FOR RESEARCH AND TEST REACTORS  

Science Conference Proceedings (OSTI)

This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

Hakan Ozaltun & Herman Shen

2011-11-01T23:59:59.000Z

254

ORNL/TM-2012/380 Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

2/380 2/380 Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program September 2012 Prepared by Cyrus Smith Randy Nanstad Robert Odette Dwight Clayton Katie Matlack Pradeep Ramuhalli Glenn Light DOCUMENT AVAILABILITY Reports produced after January 1, 1996, are generally available free via the U.S. Department of Energy (DOE) Information Bridge. Web site http://www.osti.gov/bridge Reports produced before January 1, 1996, may be purchased by members of the public from the following source. National Technical Information Service 5285 Port Royal Road Springfield, VA 22161 Telephone 703-605-6000 (1-800-553-6847) TDD 703-487-4639 Fax 703-605-6900

255

Neutronic safety parameters and transient analyses for Poland's MARIA research reactor.  

SciTech Connect

Reactor kinetic parameters, reactivity feedback coefficients, and control rod reactivity worths have been calculated for the MARIA Research Reactor (Swierk, Poland) for M6-type fuel assemblies with {sup 235}U enrichments of 80% and 19.7%. Kinetic parameters were evaluated for family-dependent effective delayed neutron fractions, decay constants, and prompt neutron lifetimes and neutron generation times. Reactivity feedback coefficients were determined for fuel Doppler coefficients, coolant (H{sub 2}O) void and temperature coefficients, and for in-core and ex-core beryllium temperature coefficients. Total and differential control rod worths and safety rod worths were calculated for each fuel type. These parameters were used to calculate generic transients for fast and slow reactivity insertions with both HEU and LEU fuels. The analyses show that the HEU and LEU cores have very similar responses to these transients.

Bretscher, M. M.; Hanan, N. A.; Matos, J. E.; Andrzejewski, K.; Kulikowska, T.

1999-09-27T23:59:59.000Z

256

RELAP5 Application to Accident Analysis of the NIST Research Reactor  

Science Conference Proceedings (OSTI)

Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

2012-03-18T23:59:59.000Z

257

DOE/EIS-0218-SA-3: Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program (November 2004)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SUPPLEMENT ANALYSIS FOR THE FOREIGN SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM NOVEMBER 2004 DOE/EIS-0218-SA-3 U.S. Department of Energy National Nuclear Security Administration Washington, DC Final Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program Final i TABLE OF CONTENTS Page 1. Introduction.............................................................................................................................................. 1 2. Background .............................................................................................................................................. 1 3. The Proposed Action ...............................................................................................................................

258

Regulatory Experiences for the Decommissioning of the Research Reactor in Korea  

Science Conference Proceedings (OSTI)

The first research reactor in Korea (KRR-1, TRIGA Mark-II) has operated since 1962, and the second one (KRR-2, TRIGA Mark-III), since 1972. Both of them were phased out in 1995 due to their lives and the operation of a new research reactor, HANARO (30 MW thermal power) operated by KAERI (Korea Atomic Energy Research Institute). After deciding the shutdown by the Nuclear Development and Utilization Committee in March 1996, KAERI began to prepare the decommissioning plan, including the environmental impact assessment, and submitted the plan to the Ministry of Science and Technology (MOST) in December 1998. Korea Institute of Nuclear Safety (KINS) reviewed document and prepared the review report in 1999. KINS is an organization of technical expertise which performs regulatory functions, entrusted by the MOST in accordance with the Atomic Energy Act and its Enforcement Decree. The review report written by KINS was consulted by the Special Committee on Nuclear Safety in January 2000. The committee submitted their consultation results to the Nuclear Safety Commission for the final approval by the Minister of MOST. The license was issued in November 2000. With the consent of the Korean government to the US Record of Decision, the spent fuel of KRR-1 and 2 was safely transported to the United States in July 1998. The decontamination and dismantling of KRR-2 was completed at the end of 2005 but the decommissioning of KRR-1 has been suspended by the problem for the memorial of the reactor. After the decommissioning of the research reactor is finished, the site will be returned to the site owner, Korea Electric Power Corporation (KEPCO). In this paper, the state-of-art and lessons learnt from recent regulatory activities for decommissioning of KRR- 2 are summarized. In conclusion: since the shutdown of KRR-1 and 2 had been decided, the safe assessment and licensing review were carried out after applying for decommissioning plan of those research reactors by operator. Through the safety assessment and license review, the plan was approved in 2000. The D and D of KRR-2 except Reactor hall of KRR-1 has been performed safely and completed in 2005. On the other hand the integrated safety for the decommissioning has been confirmed by regulatory team inspection. The radioactive waste arising from the dismantling work has been packed mainly in a 4 m{sup 3} containers and stored on site, reactor hall of KRR-2, until the LILW disposal facility is operational. For the whole dismantling work, with respect to the ALARA principle, not only worker protection from radiation and industrial hazard, but also the environment protection should be the first priority. Also much attention has been paid to the record keeping for the future decommissioning of nuclear power plants.

CHOI, Kyung-Woo [Korea Institute of Nuclear Safety, P.O Box 114, Yuseong, Daejeon (Korea, Republic of)

2008-01-15T23:59:59.000Z

259

Nuclear reactor safety. Progress report, January 1-March 31, 1982  

SciTech Connect

The work that is highlighted here represents accomplishments for the period January 1-March 31, 1982 by the groups at Los Alamos involved in reactor safety research for the Division of Accident Evaluation, Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission. Presented are brief overviews compiled by project, along with a bibliography of Technical Notes and publications written during this quarter. Information is presented concerning the TRAC code development; thermal-hydraulic analysis for PWR after ECCS operation; failure criteria for graphites used in HTGR type reactors; upper structure dynamics experiments; CRBR loss-of-flow accident analysis; and LWR severe accident analysis.

Stevenson, M.G. (comp.)

1982-08-01T23:59:59.000Z

260

Recompressed exfoliated graphite articles  

DOE Patents (OSTI)

This invention provides an electrically conductive, less anisotropic, recompressed exfoliated graphite article comprising a mixture of (a) expanded or exfoliated graphite flakes; and (b) particles of non-expandable graphite or carbon, wherein the non-expandable graphite or carbon particles are in the amount of between about 3% and about 70% by weight based on the total weight of the particles and the expanded graphite flakes combined; wherein the mixture is compressed to form the article having an apparent bulk density of from about 0.1 g/cm.sup.3 to about 2.0 g/cm.sup.3. The article exhibits a thickness-direction conductivity typically greater than 50 S/cm, more typically greater than 100 S/cm, and most typically greater than 200 S/cm. The article, when used in a thin foil or sheet form, can be a useful component in a sheet molding compound plate used as a fuel cell separator or flow field plate. The article may also be used as a current collector for a battery, supercapacitor, or any other electrochemical cell.

Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

2013-08-06T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Chemical sputtering of ATJ graphite induced by low-energy D  

E-Print Network (OSTI)

Chemical sputtering of ATJ graphite induced by low-energy Dþ 2 bombardment L.I. Vergara *, F Received 15 June 2005; accepted 4 August 2005 Abstract Results of chemical sputtering of ATJ graphite thermonuclear fusion reactor devices. Chemical and physical sputtering processes, which occur when thermal

262

Graphite and Beryllium Reflector Critical Assemblies of UO2 (Benchmark Experiments 2 and 3)  

SciTech Connect

INTRODUCTION A series of experiments was carried out in 1962-65 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for use in space reactor research programs. A core containing 93.2 wt% enriched UO2 fuel rods was used in these experiments. The first part of the experimental series consisted of 252 tightly-packed fuel rods (1.27-cm triangular pitch) with graphite reflectors [1], the second part used 252 graphite-reflected fuel rods organized in a 1.506-cm triangular-pitch array [2], and the final part of the experimental series consisted of 253 beryllium-reflected fuel rods in a 1.506-cm-triangular-pitch configuration and in a 7-tube-cluster configuration [3]. Fission rate distribution and cadmium ratio measurements were taken for all three parts of the experimental series. Reactivity coefficient measurements were taken for various materials placed in the beryllium reflected core. All three experiments in the series have been evaluated for inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP) [4] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbooks, [5]. The evaluation of the first experiment in the series was discussed at the 2011 ANS Winter meeting [6]. The evaluations of the second and third experiments are discussed below. These experiments are of interest as benchmarks because they support the validation of compact reactor designs with similar characteristics to the design parameters for a space nuclear fission surface power systems [7].

Margaret A. Marshall; John D. Bess

2012-11-01T23:59:59.000Z

263

Initiation of a phase-I trial of neutron capture therapy at the MIT research reactor  

Science Conference Proceedings (OSTI)

The Massachusetts Institute of Technology (MIT), the New England Medical Center (NEMC), and Boston University Medical Center (BUMC) initiated a phase-1 trial of boron neutron capture therapy (BNCT) on September 6, 1994, at the 5-MW(thermal) MIT research reactor (MITR). A novel form of experimental cancer therapy, BNCT is being developed for certain types of highly malignant brain tumors such as glioblastoma and melanoma. The results of the phase-1 trials on patients with tumors in the legs or feet are described.

Harling, O.K.; Bernard, J.A.; Yam, Chun-Shan [Massachussets Institute of Technology, Cambridge, MA (United States)] [and others

1995-12-31T23:59:59.000Z

264

Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices  

SciTech Connect

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

1982-03-01T23:59:59.000Z

265

Evaluation of differential shim rod worth measurements in the Oak Ridge Research Reactor  

Science Conference Proceedings (OSTI)

Reasonable agreement between calculated and measured differential shim rod worths in the Oak Ridge Research Reactor (ORR) has been achieved by taking into account the combined effects of negative reactivity contributions from changing fuel-moderator temperatures and of delayed photoneutrons. A method has been developed for extracting the asymptotic period from the shape of the initial portion of the measured time-dependent neutron flux profile following a positive reactivity insertion. In this region of the curve temperature-related reactivity feedback effects are negligibly small. Results obtained by applying this technique to differential shim rod worth measurements made in a wide variety of ORR cores are presented.

Bretscher, M.M.

1987-01-01T23:59:59.000Z

266

Diamond-graphite field emitters  

DOE Patents (OSTI)

A field emission electron emitter comprising an electrode of diamond and a conductive carbon, e.g., graphite, is provided.

Valone, Steven M. (Santa Fe, NM)

1997-01-01T23:59:59.000Z

267

Graphite-based photovoltaic cells  

DOE Patents (OSTI)

The present invention uses lithographically patterned graphite stacks as the basic building elements of an efficient and economical photovoltaic cell. The basic design of the graphite-based photovoltaic cells includes a plurality of spatially separated graphite stacks, each comprising a plurality of vertically stacked, semiconducting graphene sheets (carbon nanoribbons) bridging electrically conductive contacts.

Lagally, Max (Madison, WI); Liu, Feng (Salt Lake City, UT)

2010-12-28T23:59:59.000Z

268

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15T23:59:59.000Z

269

A disposition strategy for highly enriched, aluminum-based fuel from research and test reactors  

SciTech Connect

The strategy proposed in this paper offers the Department of Energy an approach for disposing of aluminum-based, highly enriched uranium (HEU) spent fuels from foreign and domestic research reactors. The proposal is technically, socially, and economically sound. If implemented, it would advance US non-proliferation goals while also disposing of the spent fuel`s waste by timely and proven methods using existing technologies and facilities at SRS without prolonged and controversial storage of the spent fuel. The fuel would be processed through 221-H. The radioactive fission products (waste) would be treated along with existing SRS high level waste by vitrifying it as borosilicate glass in the Defense Waste Processing Facility (DWPF) for disposal in the national geological repository. The HEU would be isotopically diluted, during processing, to low-enriched uranium (LEU) which can not be used to make weapons, thus eliminating proliferation concerns. The LEU can be sold to fabricators of either research reactor fuel or commercial power fuel. This proposed processing-LEU recycle approach has several important advantages over other alternatives, including: Lowest capital investment; lowest net total cost; quickest route to acceptable waste form and final geologic disposal; and likely lowest safety, health, and environmental impacts.

McKibben, J.M.; Gould, T.H.; McDonell, W.R.; Bickford, W.E.

1994-11-01T23:59:59.000Z

270

Thermal-hydraulic calculations for the conversion to LEU of a research reactor core  

SciTech Connect

The thermal-hydraulic analysis performed for the needs of the conversion of the open pool 5MW Greek Research Reactor (GRR-1) to a pure Low Enrichment (LEU) configuration is presented. The methodology was based on a complete set of neutronic calculations performed for the new core configuration, in compliance with pre-defined Operation Limiting Conditions. The hottest channel analysis approach was adopted, and peaking factors were used to account for fabrication or measuring uncertainties. Calculations were carried out using the numerical codes NATCON, PLTEMP and PARET provided by Argonne National Laboratory (ANL). Two main different classes of conditions were considered, namely i) steady state normal operating conditions and ii) transient cases related to accidental events including reactivity feedback effects. For steady state operating conditions the behaviour of the new configuration was examined both for forced and natural convection cooling modes. Transient calculations considered several initiating events including reactivity insertion accidents (slow or fast reactivity insertion) and total or partial loss-of-flow accidents, i.e. in accordance to guidelines provided by the IAEA for research Reactors. (author)

Grigoriadis, D. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Department of Mechanical and Manufacturing Engineering, University of Cyprus, P.O. Box 20537, Nicosia 1678 (Cyprus); Varvayanni, M.; Catsaros, N.; Stakakis, E. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece)

2008-07-15T23:59:59.000Z

271

An analysis of radionuclide behavior in water pools during accidents at the Annular Core Research Reactor  

SciTech Connect

Physical and chemical phenomena that will affect the behavior of radionuclides released from fuel in the Annular Core Research Reactor during a hypothetical, core disruptive accident are described. The phenomena include boiling of water on heated clad, metal-water reactions, vapor nucleation to form aerosol particles, coagulation of aerosol particles, aerosol deposition within bubbles rising through the shield pool, vapor dissolution in the shield pool, and revaporization of radionuclides from the shield pool. A model of these phenomena is developed and applied to predict the release of radionuclides to the confinement building of the Annular Core Research Reactor. It is found that the shield pool provides overall decontamination factors for particulate of about 2.8 {times} 10{sup 5} and decontamination factors for noble gases of about 2.5--3.7. These results are found to be sensitive to the predicted clad temperature and bubble behavior in the shield pool. Slow revalorization of krypton, xenon and iodine from the shield pool is shown to create a prolonged, low-intensity source term of radioactive material to the confinement atmosphere.

Powers, D.A.

1992-08-01T23:59:59.000Z

272

Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies  

SciTech Connect

As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.

Pond, R.B.; Matos, J.E.

1996-05-01T23:59:59.000Z

273

A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR).  

SciTech Connect

A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm{sup 3} or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study.

Hanan, N. A.

1998-01-14T23:59:59.000Z

274

Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability  

SciTech Connect

Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 C and the possible compatibility issues associated with the supercritical water environment. Reactor pressure vessel Pumps and piping

Philip E. MacDonald

2003-09-01T23:59:59.000Z

275

FUEL ASSAY REACTOR  

DOE Patents (OSTI)

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25T23:59:59.000Z

276

B Reactor | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Operational Management » History » Manhattan Project » Signature Operational Management » History » Manhattan Project » Signature Facilities » B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first large-scale plutonium production reactor. As at Oak Ridge, the need for labor turned Hanford into an atomic boomtown, with the population reaching 50,000 by summer 1944. Similar to the X-10 Graphite Reactor at Oak Ridge in terms of loading and unloading fuel, the B Reactor was built on a much larger scale and used water rather than air as a coolant. Whereas the X-10 had an initial design output of 1,000 kilowatts, the B Reactor was designed to operate at 250,000 kilowatts. Consisting of a 28- by 36-foot, 1,200-ton graphite cylinder lying on its side, the reactor was penetrated through its

277

Graphitization in C and C-Mo Steels  

Science Conference Proceedings (OSTI)

Following the recent carbon (C) and carbon-molybdenum (C-Mo) steel graphitization experience reported by several Electric Power Research Institute (EPRI) members, it became apparent that the industry could benefit from better predictive guidance to prioritize component inspections and examinations for graphitization. This research effort collected and analyzed the additional experience gained since the last EPRI project on the subject and focused on developing suitably conservative time-temperature predi...

2010-12-23T23:59:59.000Z

278

Record of Decision for the Final EIS on Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

5091 5091 Friday May 17, 1996 Part IV Department of Energy Record of Decision for the Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel; Notice 25092 Federal Register / Vol. 61, No. 97 / Friday, May 17, 1996 / Notices DEPARTMENT OF ENERGY Record of Decision for the Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel AGENCY: Department of Energy. ACTION: Record of decision. SUMMARY: DOE, in consultation with the Department of State, has decided to implement a new foreign research reactor spent fuel acceptance policy as specified in the Preferred Alternative contained in the Final Environmental Impact Statement on a Proposed

279

A neutronic feasibility study for LEU conversion of the WWR-SM research reactor in Uzbekistan.  

SciTech Connect

The WWR-SM research reactor in Uzbekistan has operated at 10 MW since 1979, using Russian-supplied IRT-3M fuel assemblies containing 90% enriched uranium. Burnup tests of three full-sized IRT-3M FA with 36% enrichment were successfully completed to a burn up of about {approximately}50% in 1987-1989. In August 1998, four IRT-3M FA with 36% enriched uranium were loaded into the core to initiate conversion of the entire core to 36% enriched fuel. This paper presents the results of equilibrium fuel cycle comparisons of the reactor using HEU (90%) and HEU (36%) IRT-3M fuel and compares results with the performance of IRT-4M FA containing LEU (19.75%). The results show that an LEU (19.75%) density of 3.8 g/cm{sup 3} is required to match the cycle length of the HEU (90%) core and an LEU density 3.9 g/cm{sup 3} is needed to match the cycle length of the HEU (36%) core.

Rakhmanov, A.

1998-10-19T23:59:59.000Z

280

A neutronic feasibility study for LEU conversion of the IR-8 research reactor.  

SciTech Connect

Equilibrium fuel cycle comparisons for the IR-8 research reactor were made for HEU(90%), HEU(36%), and LEU (19.75%) fuel assembly (FA) designs using three dimensional multi-group diffusion theory models benchmarked to detailed Monte Carlo models of the reactor. Comparisons were made of changes in reactivity, cycle length, average {sup 235}U discharge burnup, thermal neutron flux, and control rod worths for the 90% and 36% enriched IRT-3M fuel assembly and the 19.75% enriched IRT-4M fuel assembly with the same fuel management strategy. The results of these comparisons showed that a uranium density of 3.5 g/cm{sup 3} in the fuel meat would be required in the LEU IRT-4M fuel assembly to match the cycle length of the HEU(90%) IRT-3M FA and an LEU density of 3.7 g/cm{sup 3} is needed to match the cycle length of the HEU(36%) IRT-3M FA.

Deen, J. R.

1998-10-22T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

Wigner, E.P.

1958-04-22T23:59:59.000Z

282

Fuel development activities of the US RERTR Program. [Reduced Enrichment Research and Test Reactor  

SciTech Connect

Progress in the development and irradiation testing of high-density fuels for use with low-enriched uranium in research and test reactors is reported. Swelling and blister-threshold temperature data obtained from the examination of miniature fuel plates containing UAl/sub x/, U/sub 3/O/sub 8/, U/sub 3/Si/sub 2/, or U/sub 3/Si dispersed in an aluminum matrix are presented. Combined with the results of metallurgical examinations, these data show that these four fuel types will perform adequately to full burnup of the /sup 235/U contained in the low-enriched fuel. The exothermic reaction of the uranium-silicide fuels with aluminum has been found to occur at about the same temperature as the melting of the aluminum matrix and cladding and to be essentially quenched by the melting endotherm. A new series of miniature fuel plate irradiations is also discussed.

Snelgrove, J.L.; Domagala, R.F.; Wiencek, T.C.; Copeland, G.L.

1983-01-01T23:59:59.000Z

283

Transient analyses and thermal-hydraulic safety margins for the Greek Research Reactor (GRRI)  

Science Conference Proceedings (OSTI)

Various core configurations for the Greek research reactor (GRR1) have been considered in assessing the safety issues of adding a beryllium reflector to the existing water reflected HEU core and the transition from HEU to an all LEU core. The assessment has included both steady-state and transient analyses of safety margins and limits. A small all fresh Be reflected HEU core with a rather large nuclear peaking factor can still be operated safely, and thus adding a Be reflector to the larger depleted HEU core should not pose a problem. The transition mixed core with 50% LEU elements has larger void and Doppler coefficients than the HEU reference core and gives a lower peak clad temperature under transient conditions. The transition cores should give ever increasing margins to plate melting and fission product release as LEU elements are added to the core.

Woodruff, W.L.; Deen, J.R. [Argonne National Lab., IL (United States); Papastergiou, C. [National Centre for Scientific Research Demokritos, Athens (Greece)

1995-02-01T23:59:59.000Z

284

Reactor pressure vessel integrity research at the Oak Ridge National Laboratory  

Science Conference Proceedings (OSTI)

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program.

Corwin, W.R.; Pennell, W.E.; Pace, J.V.

1995-12-31T23:59:59.000Z

285

Neutronic safety and transient analyses for potential LEU conversion of the IR-8 research reactor.  

SciTech Connect

Kinetic parameters, isothermal reactivity feedback coefficients and three transients for the IR-8 research reactor cores loaded with either HEU(90%), HEU(36%), or LEU (19.75%) fuel assemblies (FA) were calculated using three dimensional diffusion theory flux solutions, RELAP5/MOD3.2 and PARET. The prompt neutron generation time and effective delayed neutron fractions were calculated for fresh and beginning-of-equilibrium-cycle cores. Isothermal reactivity feedback coefficients were calculated for changes in coolant density, coolant temperature and fuel temperature in fresh and equilibrium cores. These kinetic parameters and reactivity coefficients were used in transient analysis models to predict power histories, and peak fuel, clad and coolant temperatures. The transients modeled were a rapid and slow loss-of-flow, a slow reactivity insertion, and a fast reactivity insertion.

Deen, J. R.; Hanan, N. A.; Smith, R. S.; Matos, J. E.; Egorenkov, P. M.; Nasonov, V. A.

1999-09-27T23:59:59.000Z

286

Preliminary neutronics calculations for conversion of the Tehran research reactor core from HEU to LEU fuel  

SciTech Connect

The 5-MW highly enriched uranium (HEU)-fueled Tehran Research Reactor is considered for conversion to high-density, low-enriched uranium (LEU) fuel. A preliminary neutronics calculation is performed as part of the conversion goal. In this study, two cores are considered: the HEU reference core and a proposed LEU core similar to the reference core, and a proposed LEU core similar to the reference core, using standardized U[sub 3]Si[sub 2] plates with the option of different [sup 235]U loadings. Various possibilities are investigated for the conversion of HEU to LEU fuel elements with 20% enriched [sup 235]U loadings of 207 to 297 g [sup 235]U/element. For the same equilibrium cycle length, the fuels are compared for flux, power distribution, burnup, and reactivity.

Nejat, S.M.R. (McMaster Univ., Hamilton, Ontario (Canada). Dept. of Engineering Physics.)

1993-08-01T23:59:59.000Z

287

Characterization of wastes in and around early reactors at the Hanford Site: The use of historical research  

SciTech Connect

This paper will present the waste characterization knowledge that has been gained in the first, ``Large-Scale Remediation Study`` to be performed on the reactor areas (100 Areas) of the Hanford Site. Undertaken throughout the past year, this research project has identified thousands of pieces of buried hardware, as well as the volumes of liquid wastes in burial sites in the reactor areas. The author of this landmark study, Dr. Michele Gerber, will discuss historical research as a safe and cost-effective characterization tool.

Gerber, M.S.

1993-10-01T23:59:59.000Z

288

C-1 coproduct development for N Reactor: Excerpts from N-Reactor Department research and development budget for FY-1967 and revision of budget for FY-1966  

SciTech Connect

This report discusses the production of weapon grade plutonium and tritium as an economically attractive mode of operation for N-Reactor. With the successful completion of the proposed research and development program coproduct production in N-Reactor could begin in January 1967. Approximately 5 kg of tritium could be produced at a reduction in unit total cost of up to $7 per gram equivalent when compared to other modes of operation of N-Reactor that yield only-weapon grade plutonium. Achieving this production level and unit cost requires a fuel and target design that will subject the materials to more severe temperature and exposure conditions than for the tube-in-tube design. Analyses of existing knowledge and experimental results obtained to date are very encouraging. However, there is a continuing need for additional experimental verification of the limiting conditions for both fuel and target to permit the specification of a production fuel and target design. The incremental cost of this program is estimated to be $1,055 million in FY-1965, $0.796 million in FY-1966 and $0.450 million in FY-1967 for a total of $2.3 million. The principal benefits to be derived from this expenditure of research and development funds are lower unit grade plutonium. When the reactor reaches equilibrium operation as a costs for both tritium and weapon dual-purpose reactor producing process steam for credit, it will be a highly competitive producer of plutonium and tritium in the AEC complex.

1965-04-19T23:59:59.000Z

289

Heat exchanger using graphite foam  

DOE Patents (OSTI)

A heat exchanger is disclosed. The heat exchanger may have an inlet configured to receive a first fluid and an outlet configured to discharge the first fluid. The heat exchanger may further have at least one passageway configured to conduct the first fluid from the inlet to the outlet. The at least one passageway may be composed of a graphite foam and a layer of graphite material on the exterior of the graphite foam. The layer of graphite material may form at least a partial barrier between the first fluid and a second fluid external to the at least one passageway.

Campagna, Michael Joseph; Callas, James John

2012-09-25T23:59:59.000Z

290

Lithium Diffusion in Graphitic Carbon  

NLE Websites -- All DOE Office Websites (Extended Search)

Volume 1 Start Page 1176 Issue 8 Pagination 1176-1180 Keywords anode, diffusion, graphene, lithium ion battery, transport Abstract Graphitic carbon is currently considered the...

291

Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.  

SciTech Connect

This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.

Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

2008-06-23T23:59:59.000Z

292

Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one-on-one interviews and traditional and online focus groups with scientists. The latter include SNS, HFIR, and APS users as well as scientists at ORNL, some of whom had not yet used HFIR and/or SNS. These approaches informed development of the second phase, a quantitative online survey. The survey consisted of 16 questions and 7 demographic categorizations, 9 open-ended queries, and 153 pre-coded variables and took an average time of 18 minutes to complete. The survey was sent to 589 SNS/HFIR users, 1,819 NSLS users, and 2,587 APS users. A total of 899 individuals provided responses for this study: 240 from NSLS; 136 from SNS/HFIR; and 523 from APS. The overall response rate was 18%.

Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

2011-03-01T23:59:59.000Z

293

Optimization of irradiation conditions for {sup 177}Lu production at the LVR-15 research reactor  

Science Conference Proceedings (OSTI)

The use of lutetium in medicine has been increasing over the last few years. The {sup 177}Lu radionuclide is commercially available for research and test purposes as a diagnostic and radiotherapy agent in the treatment of several malignant tumours. The yield of {sup 177}Lu from the {sup 176}Lu(n,{gamma}){sup 177}Lu nuclear reaction depends significantly on the thermal neutron fluence rate. The capture cross-sections of both reaction {sup 176}Lu(n,{gamma}){sup 177}Lu and reaction {sup 177}Lu(n,{gamma}){sup 178}Lu are very high. Therefore a burn-up of target and product nuclides should be taken into account when calculating {sup 177}Lu activity. The maximum irradiation time, when the activity of the {sup 177}Lu radionuclide begins to decline, was found for different fluence rates. Two vertical irradiation channels at the LVR-15 nuclear research reactor were compared in order to choose the channel with better irradiation conditions, such as a higher thermal neutron fluence rate in the irradiation volume. In this experiment, lutetium was irradiated in a titanium capsule. The influence of the Ti capsule on the neutron spectrum was monitored using activation detectors. The choice of detectors was based on requirements for irradiation time and accurate determination of thermal neutrons. The following activation detectors were selected for measurement of the neutron spectrum: Ti, Fe, Ni, Co, Ag and W. (authors)

Lahodova, Z.; Viererbl, L.; Klupak, V. [Research Centre Rez Ltd., Husinec-130, Rez 250 67 (Czech Republic); Srank, J. [Nuclear Physics Inst. of the Academy of Sciences, Husinec-130, Rez 250 67 (Czech Republic)

2012-07-01T23:59:59.000Z

294

Final Site Specific Decommissioning Inspection Report #2 for the University of Washington Research and Test Reactor, Seattle, Washington  

SciTech Connect

During the period of August through November 2006, ORISE performed a comprehensive IV at the University of Washington Research and Test Reactor Facility. The objective of the ORISE IV was to validate the licensees final status survey processes and data, and to assure the requirements of the DP and FSSP were met.

S.J. Roberts

2007-03-20T23:59:59.000Z

295

Comparison of shell-and-tube with frame-and-plate-type heat exchangers for the MIT research reactor upgrade  

SciTech Connect

This paper discusses a comparison of shell-and-tube with frame-and-plate-type heat exchangers for the proposed upgrade of the Massachusetts Institute of Technology research reactor (MITR). The comparison is based on the following considerations: thermal-hydraulic performance, maintenance, personnel dose rate, and pricing.

Ser, Yorick, Hu, Lin-Wen [Massachusetts Institute of Technology, Cambridge, MA (United States)

1997-12-01T23:59:59.000Z

296

Analysis of in-core experiment activities for the MIT Research Reactor using the ORIGEN computer code  

E-Print Network (OSTI)

The objective of this study is to devise a method for utilizing the ORIGEN-S computer code to calculate the activation products generated in in-core experimental assemblies at the MIT Research Reactor (MITR-II). ORIGEN-S ...

Helvenston, Edward M. (Edward March)

2006-01-01T23:59:59.000Z

297

Reactor safety research programs. Quarterly report, July 1-September 30, 1979  

SciTech Connect

The programs include: ultimate heat sink performance measurement; experimental verification of steady state codes: Task A - irradiation results and Task C - code development; graphite nondestructive testing; acoustic emission-flaw relationship for in-service monitoring of nuclear pressure vessels; fuel subassembly procurement and irradiation test program; report of resident engineer at Cadarache, France; core thermal model development; integration of nondestructive examination reliability and fracture mechanics; and steam generator tube integrity.

Hooper, J.L. (comp.)

1980-03-01T23:59:59.000Z

298

Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 5: Graphite PIRTs  

DOE Green Energy (OSTI)

Here we report the outcome of the application of the Nuclear Regulatory Commission (NRC) Phenomena Identification and Ranking Table (PIRT) process to the issue of nuclear-grade graphite for the moderator and structural components of a next generation nuclear plant (NGNP), considering both routine (normal operation) and postulated accident conditions for the NGNP. The NGNP is assumed to be a modular high-temperature gas-cooled reactor (HTGR), either a gas-turbine modular helium reactor (GTMHR) version [a prismatic-core modular reactor (PMR)] or a pebble-bed modular reactor (PBMR) version [a pebble bed reactor (PBR)] design, with either a direct- or indirect-cycle gas turbine (Brayton cycle) system for electric power production, and an indirect-cycle component for hydrogen production. NGNP design options with a high-pressure steam generator (Rankine cycle) in the primary loop are not considered in this PIRT. This graphite PIRT was conducted in parallel with four other NRC PIRT activities, taking advantage of the relationships and overlaps in subject matter. The graphite PIRT panel identified numerous phenomena, five of which were ranked high importance-low knowledge. A further nine were ranked with high importance and medium knowledge rank. Two phenomena were ranked with medium importance and low knowledge, and a further 14 were ranked medium importance and medium knowledge rank. The last 12 phenomena were ranked with low importance and high knowledge rank (or similar combinations suggesting they have low priority). The ranking/scoring rationale for the reported graphite phenomena is discussed. Much has been learned about the behavior of graphite in reactor environments in the 60-plus years since the first graphite rectors went into service. The extensive list of references in the Bibliography is plainly testament to this fact. Our current knowledge base is well developed. Although data are lacking for the specific grades being considered for Generation IV (Gen IV) concepts, such as the NGNP, it is fully expected that the behavior of these graphites will conform to the recognized trends for near isotropic nuclear graphite. Thus, much of the data needed is confirmatory in nature. Theories that can explain graphite behavior have been postulated and, in many cases, shown to represent experimental data well. However, these theories need to be tested against data for the new graphites and extended to higher neutron doses and temperatures pertinent to the new Gen IV reactor concepts. It is anticipated that current and planned future graphite irradiation experiments will provide the data needed to validate many of the currently accepted models, as well as providing the needed data for design confirmation.

Burchell, Timothy D [ORNL; Bratton, Rob [Idaho National Laboratory (INL); Marsden, Barry [University of Manchester, UK; Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission; Penfield, Scott [Technology Insights; Mitchell, Mark [PBMR (Pty) Ltd.; Windes, Will [Idaho National Laboratory (INL)

2008-03-01T23:59:59.000Z

299

Graphitic packing removal tool  

DOE Patents (OSTI)

Graphitic packing removal tools are described for removal of the seal rings in one piece from valves and pumps. The packing removal tool has a cylindrical base ring the same size as the packing ring with a surface finish, perforations, knurling or threads for adhesion to the seal ring. Elongated leg shanks are mounted axially along the circumferential center. A slit or slits permit insertion around shafts. A removal tool follower stabilizes the upper portion of the legs to allow a spanner wrench to be used for insertion and removal.

Meyers, K.E.; Kolsun, G.J.

1996-12-31T23:59:59.000Z

300

Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents  

Science Conference Proceedings (OSTI)

The US Department of Energy is performing research and development (R&D) that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP) Program / GEN-IV Very High Temperature Reactor (VHTR). Phenomena identification and ranking studies (PIRT) to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Schultz et al., 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) are very high priority for the NGNP program. Following a loss of coolant and system depressurization, air will enter the core through the break. Air ingress leads to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heat-up of the bottom reflector and the reactor core and will cause the release of fission products eventually. The potential collapse of the bottom reflector because of burn-off and the release of CO lead to serious safety problems. For estimation of the proper safety margin we need experimental data and tools, including accurate multi-dimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model. We also need to develop effective strategies to mitigate the effects of oxidation. The results from this research will provide crucial inputs to the INL NGNP/VHTR Methods R&D project. This project is focused on (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the bottom reflector, (d) structural tests of the burnt-off bottom reflector, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i) verification and validation of the coupled models.

Chang Ho Oh; Eung Soo Kim; Hee Cheon No; Nam Zin Cho

2008-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor  

DOE Green Energy (OSTI)

The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

2009-10-01T23:59:59.000Z

302

License amendment for neutron capture therapy at the MIT research reactor  

SciTech Connect

This paper reports the issuance by the U.S. Nuclear Regulatory Commission (NRC) of a license amendment to the Massachusetts Institute of Technology (MIT) for the use of the MIT Research Reactor's (MITR-II) medical therapy facility beam for the treatment of humans using neutron capture therapy (NCT). This amendment is one of 11 required approvals. The others are those of internal MIT committees, review panels of the Tufts-New England Medical Center (NEMC), which is directing the program jointly with MIT, that of the U.S. Food and Drug Administration, and an NRC amendment to the NEMC hospital license. This amendment is the first of its type to be issued by NRC, and as such it establishes a precedent for the conduct of human therapy using neutron beams. Neutron capture therapy is a bimodal method for treating cancer that entails the administration of a tumor-seeking boronated drug followed by the irradiation of the target organ with neutrons. The latter cause boron nuclei to fission and thereby release densely ionizing helium and lithium nuclei, which destroy cancerous cells while leaving adjacent healthy cells undamaged. Neutron capture therapy is applicable to glioblastoma multiforme (brain tumors) and metastasized melanoma (skin cancer). Both Brookhaven National Laboratory and MIT conducted trials of NCT more than 30 yr ago. These were unsuccessful because the available boron drugs did not concentrate sufficiently in tumor and because the thermal neutron beams that were used did not enable neutrons to travel deep enough into the brain.

Bernard, J.A. (Massachusetts Institute of Technology, Cambride, MA (United States))

1993-01-01T23:59:59.000Z

303

Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel  

SciTech Connect

This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

1994-10-01T23:59:59.000Z

304

Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 20092013  

SciTech Connect

Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to declineeven with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement and leadership on nuclear safety and security issues.

Idaho National Laboratory

2009-12-01T23:59:59.000Z

305

Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009201/span>3  

Science Conference Proceedings (OSTI)

Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to declineeven with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement and leadership on nuclear safety and security issues.

Idaho National Laboratory

2009-12-01T23:59:59.000Z

306

Relation between measurable and principal characteristics of radiation-induced shape-change of graphite  

E-Print Network (OSTI)

On the basis of studies of radiation-induced shape-change of reactor graphite GR-280, through the series of measurements of samples with different orientation of cutting with respect to the direction of extrusion, a conclusion is made about the existence of polycrystal substructural elements - domains. Domains, like graphite as a whole, possess the property of transverse isotropy, but have different amplitudes of shape-change and random orientations of the axes of axial symmetry. The model of graphite, constructed on the basis of the concept of domains allowed to explain from a unified point of view most of existing experimental data. It is shown that the presence of the disoriented domain structure leads to the development of radiation-induced stresses and to the dependence of the shape-change on the size of graphite samples. We derive the relation between the shape-change of finite size samples and the actual shape-change of macro-graphite.

M. V. Arjakov; A. V. Subbotin; S. V. Panyukov; O. V. Ivanov; A. S. Pokrovskii; D. V. Kharkov

2011-04-12T23:59:59.000Z

308

Testing of Small Graphite Samples for Nuclear Qualification  

Science Conference Proceedings (OSTI)

Accurately determining the mechanical properties of small irradiated samples is crucial to predicting the behavior of the overal irradiated graphite components within a Very High Temperature Reactor. The sample size allowed in a material test reactor, however, is limited, and this poses some difficulties with respect to mechanical testing. In the case of graphite with a larger grain size, a small sample may exhibit characteristics not representative of the bulk material, leading to inaccuracies in the data. A study to determine a potential size effect on the tensile strength was pursued under the Next Generation Nuclear Plant program. It focuses first on optimizing the tensile testing procedure identified in the American Society for Testing and Materials (ASTM) Standard C 781-08. Once the testing procedure was verified, a size effect was assessed by gradually reducing the diameter of the specimens. By monitoring the material response, a size effect was successfully identified.

Julie Chapman

2010-11-01T23:59:59.000Z

309

Inspection methods for physical protection. Task II. Review of research reactor licensees' physical security practices  

SciTech Connect

Security systems and security procedures for the AFRRI reactor, the University of Maryland TRIGA reactor, and the University of Virginia CAVALIER and UVAR reactors are described.

1980-07-03T23:59:59.000Z

310

Russian RBMK reactor design information  

Science Conference Proceedings (OSTI)

This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received.

Not Available

1993-11-01T23:59:59.000Z

311

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion  

SciTech Connect

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

2012-09-30T23:59:59.000Z

312

Novel, Magnetically Fluidized-Bed Reactor Development for the Looping Process: Coal to Hydrogen Production Research and Development  

NLE Websites -- All DOE Office Websites (Extended Search)

Novel, Magnetically Fluidized-Bed Novel, Magnetically Fluidized-Bed Reactor Development for the Looping Process: Coal to Hydrogen Production Research and Development Background The U.S. Department of Energy (DOE) National Energy Technology Laboratory (NETL) is committed to improving methods for co-producing power and chemicals, fuels, and hydrogen (H2). Gasification is a process by which fuels such as coal can be used to produce synthesis gas (syngas), a mixture of H2, carbon monoxide (CO), and carbon

313

Using Graphite to view network data  

NLE Websites -- All DOE Office Websites (Extended Search)

Graphite Graphite to Visualize Network Data Jon Dugan Summer ESCC 2010, Columbus, OH Lawrence Berkeley National Laboratory U.S. Department of Energy | Office of Science ESnet Statistics Overview ESxSNMP (Data Collection) ESxSNMP (Data Collection) Graphite (Visualization) Graphite (Visualization) Analytics (Custom Reports) Analytics (Custom Reports) Net Almanac (Metadata) Net Almanac (Metadata) Lawrence Berkeley National Laboratory U.S. Department of Energy | Office of Science What is Graphite? "Enterprise scalable realtime graphing" * Developed by Orbitz for visualizing internal performance data * Open source: https://launchpad.net/graphite * Has own RRD like database called Carbon * RRD Compatible ESxSNMP Integration * via REST interface * Easy integration, Graphite is well written

314

Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor  

E-Print Network (OSTI)

Molten salt is a promising coolant candidate for Advanced High Temperature Reactor (AHTR) Gen-IV designs. The low neutron absorption, high thermal capacity, chemical inertness, and high boiling point at low pressure of ...

Bean, Malcolm K.

2011-08-01T23:59:59.000Z

315

Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior  

SciTech Connect

This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

Monteleone, S. [comp.

1995-04-01T23:59:59.000Z

316

Status of Initial Assessment of Physical and Mechanical Properties of Graphite Grades for NGNP Appkications  

Science Conference Proceedings (OSTI)

Current candidate graphite grades for the core structures of NGNP include grades NBG-17, NBG-18, PCEA and IG-430. Both NBG-17 and NBG-18 are manufactured using pitch coke, and are vibrationally molded. These medium grain products are produced by SGL Carbon SAS (France). Tayo Tanso (Japan) produces IG-430 which is a petroleum coke, isostatically molded, nuclear grade graphite. And PCEA is a medium grain, extruded graphite produced by UCAR Carbon Co. (USA) from petroleum coke. An experimental program has been initiated to develop physical and mechanical properties data for these current candidate graphites. The results will be judged against the requirements for nuclear grade graphites set forth in ASTM standard D 7219-05 "Standard Specification for Isotropic and Near-isotropic Nuclear Graphites". Physical properties data including thermal conductivity and coefficient of thermal expansion, and mechanical properties data including tensile, compressive and flexural strengths will be obtained using the established test methods covered in D-7219 and ASTM C 781-02 "Standard Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear Reactors". Various factors known to effect the properties of graphites will be investigated. These include specimen size, spatial location within a graphite billet, specimen orientation (ag and wg) within a billet, and billet-to-billet variations. The current status of the materials characterization program is reported herein. To date billets of the four graphite grades have been procured, and detailed cut up plans for obtaining the various specimens have been prepared. Particular attention has been given to the traceability of each specimen to its spatial location and orientation within a billet.

Strizak, Joe P [ORNL; Burchell, Timothy D [ORNL; Windes, Will [Idaho National Laboratory (INL)

2011-12-01T23:59:59.000Z

317

The Effect of Neutron Irradiation on the Fracture Toughness of Graphite  

SciTech Connect

As part of our irradiated graphite recycle program a small quantity of PCEA grade graphite was irradiated in the High Flux Isotope Reactor (HFIR) at ORNL. The graphite will provide the raw material for future recycle experiments. The geometry of the irradiated graphite allowed us to study the effects of neutron irradiation on the Critical Stress Intensity Factor, KIc, of graphite. The specimens where irradiated in two groups of 6 at an irradiation temperature of 900 C in rabbit capsules to doses of 6.6 and 10.2 DPA, respectively. Following a full suite of pre-and post-irradiation examination, which included dimensions, mass, electrical resistivity, elastic constants, and thermal expansion (to 800 C) the samples were notched and tested to determine their KIc using the newly approved ATSM test method for SENB fracture toughness of graphite. Here we report the irradiation induced changes in the dimensions, elastic constants, resistivity, and coefficient of thermal expansion of PCEA graphite. Moreover, irradiation induced changes in the Critical Stress Intensity Factor, KIc, or fracture toughness, are reported and discussed. Very little work on the effect of neutron irradiation on the fracture toughness of graphite has previously be performed or reported.

Burchell, Timothy D [ORNL; Strizak, Joe P [ORNL

2012-01-01T23:59:59.000Z

318

Development of refined MCNPX-PARET multi-channel model for transient analysis in research reactors  

SciTech Connect

Reactivity insertion transients are often analyzed (RELAP, PARET) using a two-channel model, representing the hot assembly with specified power distribution and an average assembly representing the remainder of the core. For the analysis of protected by the reactor safety system transients and zero reactivity feedback coefficients this approximation proves to give adequate results. However, a more refined multi-channel model representing the various assemblies, coupled through the reactivity feedback effects to the whole reactor core is needed for the analysis of unprotected transients with excluded over power and period trips. In the present paper a detailed multi-channel PARET model has been developed which describes the reactor core in different clusters representing typical BR2 fuel assemblies. The distribution of power and reactivity feedback in each cluster of the reactor core is obtained from a best-estimate MCNPX calculation using the whole core geometry model of the BR2 reactor. The sensitivity of the reactor response to power, temperature and energy distributions is studied for protected and unprotected reactivity insertion transients, with zero and non-zero reactivity feedback coefficients. The detailed multi-channel model is compared vs. simplified fewer-channel models. The sensitivities of transient characteristics derived from the different models are tested on a few reactivity insertion transients with reactivity feedback from coolant temperature and density change. (authors)

Kalcheva, S.; Koonen, E. [SCK-CEN, BR2 Reactor Dept., Boeretang 200, 2400 Mol (Belgium); Olson, A. P. [RERTR Program, Nuclear Engineering Div., Argonne National Laboratory, Cass Avenue, Argonne, IL 60439 (United States)

2012-07-01T23:59:59.000Z

319

J. Plasma Fusion Res. SERIES, Vol. 10 (2013) Flibe-Tritium Research for Fission or Fusion Reactors at Kyushu University  

E-Print Network (OSTI)

There is increasing interest in using ionic molten-salt Flibe not only as self-cooled tritium(T)-breeding material in a fusion reactor blanket but also as fuel solvent of molten-salt fission reactors. Application of Flibe to T-breeding fluid for a stellarator-type fusion reactor operated at a high magnetic field brings large simplification of its blanket structure, allowing continuous operation under high-beta plasma conditions. Using mixed Flibe-ThF 4+UF 4 fuel in molten salt fission reactors permits stable long-term operation without fuel exchange. When Flibe or Flinak is irradiated by neutrons, however, acid and corrosive TF is generated, and some T permeates through structural walls. In order to solve these problems, chemical conditions of Flibe are changed using the redox-control reaction, Be+2TF=BeF 2+T 2. In addition, permeation of hydrogen isotopes is lowered by enhancing T recovery rates. Part of Flibe-tritium researches are performed at Idaho National Laboratory (INL) under the Japan-US collaboration work of JUPITER-II. Our own contributions to the topics are shortly introduced in this paper.

Satoshi Fukada

2012-01-01T23:59:59.000Z

320

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core  

Science Conference Proceedings (OSTI)

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried out at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.

Bokhari, Ishtiaq H. [Pakistan Institute of Nuclear Science and Technology (Pakistan)

2004-12-15T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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321

40-MW(E) PROTOTYPE HIGH-TEMPERATURE GAS-COOLED REACTOR RESEARCH AND DEVELOPMENT PROGRAM. Summary Report for the Period January 1, 1959-December 31, 1959 and Quarterly Progress Report for the Period October 1, 1959-December 31, 1959  

SciTech Connect

The HTGR prototype plant (Peach Bottom Power Reactor) is being designed to produce steam at l450 psi and 1000 deg F and to have a net capacity of 40 Mw(e). The fuel temperatures and gas pressures will be approximately the same as those required for larger plants. The reactor data and operating conditions for the graphite-clad core are given. The reactor and primary coolant systems are described. The prospects for development of the graphite-clad fuel element in time for use in the first loading of the reactor were improved by important advances in methods of fabrication and testing of both fuel compacts and graphite sleeves. The hot-pressing process for making fuel compacts was used successfully to make full-size compacts with a uniform distribution of ThC/sub 2/- UC/sub 2/ particles. Three irradiation capsules were fabricated and inserted in a test reactor to determine fuel compact and sleeve performance under HTGR conditions of irradiation and temperature. Two of these ran satisfactorily for the scheduled time of operation. A scope design study of the in-pile loop that will be used to evaluate the full-diameter graphite-clad element was completed. Experiments to determine the extent of fuel migration within the element were undertaken. Preliminary results indicated that the central fuel-element temperatures must not exceed 2300-C for routine operation. An important start was made in developing an understanding of how to treat the neutron thermalization process in high-temperature graphite reactors. Analytical techniques for calculating the thermal neutron spectra in poisoned graphite media were developed and programmed for the IBM 704 computer. The experimental technique of measuring neutron spectra by using a pulsed linear electron accelerator was demonstrated by measurements made with boron-loaded graphite. A mockup of a small portion of the reactor core was constructed and operated to determine the local heat-transfer coefficients and pressure drop in the tricusp- shaped coolant passage. Initial results indicated that the variation of the heat-transfer coefficient around the circumference of the element is less than expected. Studies were started of the transient temperatures and stresses developed in the pressure vessel as a result of load changes or a scram. A detailed study of several types of steam generator for use in the nuclear steam supply system was completed. A design incorporating a steam drum was selected for further study. Preliminary flow diagrams were completed for the helium- purification and fission-product trapping systems. Adsorption isobars for selected fission products in activated carbon were measured and will be used in the detailed design of the trapping system. Detailed planning of the experimental reactor physics program was initiated. Progress was made in the identification of the principal safeguards problems for this type of reactor, and a preliminary safety analysis of the plant was completed. (auth)

1960-09-01T23:59:59.000Z

322

Expanding and optimizing fuel management and data analysis capabilities of MCODE-FM in support of MIT research reactor (MITR-II) LEU conversion  

E-Print Network (OSTI)

Studies are underway in support of the MIT research reactor (MITR-II) conversion from high enriched Uranium (HEU) to low enriched Uranium (LEU), as required by recent non-proliferation policy. With the same core configuration ...

Horelik, Nicholas E. (Nicholas Edward)

2012-01-01T23:59:59.000Z

323

Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II)  

E-Print Network (OSTI)

The MIT Research Reactor (MITR-II) is currently undergoing analysis for the planned conversion from high enriched uranium (HEU) to low enriched uranium (LEU), as part of a global effort to minimize the availability of ...

Connaway, Heather M. (Heather Moira)

2012-01-01T23:59:59.000Z

324

PROCEEDINGS OF THE AEC SYMPOSIUM FOR CHEMICAL PROCESSING OF IRRADIATED FUELS FROM POWER, TEST, AND RESEARCH REACTORS, RICHLAND, WASHINGTON, OCTOBER 20 AND 21, 1959  

SciTech Connect

A review is presented in this symposium of the technology currently available for processing spent fuels from research, test, and power reactors. Twenty-one papers are included. Separate abstracts have been prepared for each paper. (W.L.H.)

1960-01-01T23:59:59.000Z

325

PRELIMINARY HAZARDS SUMMARY REPORT ON THE BROOKHAVEN HIGH FLUX BEAM RESEARCH REACTOR  

SciTech Connect

The High Flux Beam Reactor, HFBR, is cooled, moderated, and reflected by heavy water and designed to produce 40 Mw with a total epithermal flux of ~1.6 X 10/sup 15/cm/sup -2/ sec/sup -1/ and a flector thermal maximum flux of 7 X 10/sup 14/ cm/sup -2/ sec/sup -1/, using a core formed by ETR plate-type fuel elements in a close-packed array. The hazards summary is given in terms of site description, reactor design, building design, plant operation, disposal of radioactive wastes and effluents, and safety analysis. (B.O.G.)

Hendrie, J.M.; Kouts, H.J.C.

1961-05-01T23:59:59.000Z

326

EXPERIMENT OPERATIONS PLAN FOR A LOSS-OF-COOLANT ACCIDENT SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR  

SciTech Connect

Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship between the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. This document contains both experiment proposal and assembly proposal information. The intent of this document is to supply information required by the Chalk River Nuclear Laboratories (CRNL), and to identify the planned procedures and data that will be used both to establish readiness to proceed from one test phase to the next and to operate the experiment. Operating control settings and limits are provided for both experimenter systems and CRNL systems. A hazards review summarizes safety issues that have been addressed during the development of the experiment plan.

Russcher, G. E.; Cannon, L. W.; Goodman, R. L.; Hesson, G. M.; King, L. L.; McDuffie, P. N.; Marshall, R. K.; Nealley, C.; Pilger, J. P.; Mohr, C. L.

1981-04-01T23:59:59.000Z

327

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Complex Description Complex Description Current HFBR Complex The HFBR complex consists of multiple structures and systems that were necessary to operate and maintain the reactor. The most recognizable features of the complex are the domed reactor confinement building and the distinctive red-and-white stack. Portions of the complex building structures, systems, and components, some of which are underground, were contaminated with radionuclides and chemicals as a result of previous HFBR and Brookhaven Graphite Research Reactor (BGRR) operations. A number of decommissioning and preparation for long-term safe storage actions have been taken including the removal of contaminated structures, hazardous materials, and contaminated equipment and components. The structures and systems, both current and former, are

328

MIT Modular Pebble Bed Reactor (MPBR) A Summary of Research Activities and Accomplishments  

E-Print Network (OSTI)

operation #12;MIT MPBR Specifications Thermal Power 250 MW - 120 Mwe Target Thermal Efficiency 45 % Core for a pebble bed reactor power plant system with high efficiency and minimum capital cost ­ Net efficiency > 45;Plant With Space Frames #12;#12;For 1150 MW Electric Power Station Turbine Hall Boundary Admin Training

329

Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.  

Science Conference Proceedings (OSTI)

This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design

Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

2007-06-30T23:59:59.000Z

330

Examples of the use of PSA in the design process and to support modifications at two research reactors  

Science Conference Proceedings (OSTI)

Many, if not most, of the world`s commercial nuclear power plants have been the subject of plant-specific probabilistic safety assessments (PSA). A growing number of other nuclear facilities as well as other types of industrial installations have been the focus of plant-specific PSAs. Such studies have provided valuable information concerning the nature of the risk of the individual facility and have been used to identify opportunities to manage that risk. This paper explores the risk management activities associated with two research reactors in the United States as a demonstration of the versatility of the use of PSA to support risk-related decision making.

Johnson, D.H.; Bley, D.C.; Lin, J.C. [PLG, Inc., Newport Beach, CA (United States); Ramsey, C.T.; Linn, M.A. [Oak Ridge National Lab., TN (United States)

1994-03-01T23:59:59.000Z

331

Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics  

SciTech Connect

This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1993-03-01T23:59:59.000Z

332

Fluid-Structure Interaction for Coolant Flow in Research-type Nuclear Reactors  

Science Conference Proceedings (OSTI)

The High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), is scheduled to undergo a conversion of the fuel used and this proposed change requires an extensive analysis of the flow through the reactor core. The core consists of 540 very thin and long fuel plates through which the coolant (water) flows at a very high rate. Therefore, the design and the flow conditions make the plates prone to dynamic and static deflections, which may result in flow blockage and structural failure which in turn may cause core damage. To investigate the coolant flow between fuel plates and associated structural deflections, the Fluid-Structure Interaction (FSI) module in COMSOL will be used. Flow induced flutter and static deflections will be examined. To verify the FSI module, a test case of a cylinder in crossflow, with vortex induced vibrations was performed and validated.

Curtis, Franklin G [ORNL; Ekici, Kivanc [ORNL; Freels, James D [ORNL

2011-01-01T23:59:59.000Z

333

First Direct Evidence of Dirac Fermions in Graphite  

NLE Websites -- All DOE Office Websites (Extended Search)

Direct Evidence of Dirac Direct Evidence of Dirac Fermions in Graphite First Direct Evidence of Dirac Fermions in Graphite Print Wednesday, 27 June 2007 00:00 The recent surge of interest in the electronic properties of graphene-that is, isolated layers of graphite just one atomic layer thick-has largely been driven by the discovery that electron mobility in graphene is ten times higher than in commercial-grade silicon, raising the possibility of high-efficiency, low-power, carbon-based electronics. Scientists attribute graphene's surprising current capacity (as well as a number of even stranger phenomena) to the presence of charge carriers that behave as if they are massless, "relativistic" quasiparticles called Dirac fermions. Harnessing these quasiparticles in real-world carbon-based devices, however, requires a deeper knowledge of their behavior under less-than-ideal circumstances, such as around defects, at edges, or in three dimensions-in other words, in graphite. At the ALS, a team of researchers using angle-resolved photoemission spectroscopy (ARPES) have now produced the first direct evidence of massless Dirac fermions in graphite coexisting with quasiparticles of finite effective mass and defect-induced localized states.

334

First Direct Evidence of Dirac Fermions in Graphite  

NLE Websites -- All DOE Office Websites (Extended Search)

Direct Evidence of Dirac Fermions in Graphite Print Direct Evidence of Dirac Fermions in Graphite Print The recent surge of interest in the electronic properties of graphene-that is, isolated layers of graphite just one atomic layer thick-has largely been driven by the discovery that electron mobility in graphene is ten times higher than in commercial-grade silicon, raising the possibility of high-efficiency, low-power, carbon-based electronics. Scientists attribute graphene's surprising current capacity (as well as a number of even stranger phenomena) to the presence of charge carriers that behave as if they are massless, "relativistic" quasiparticles called Dirac fermions. Harnessing these quasiparticles in real-world carbon-based devices, however, requires a deeper knowledge of their behavior under less-than-ideal circumstances, such as around defects, at edges, or in three dimensions-in other words, in graphite. At the ALS, a team of researchers using angle-resolved photoemission spectroscopy (ARPES) have now produced the first direct evidence of massless Dirac fermions in graphite coexisting with quasiparticles of finite effective mass and defect-induced localized states.

335

First Direct Evidence of Dirac Fermions in Graphite  

NLE Websites -- All DOE Office Websites (Extended Search)

Direct Evidence of Dirac Fermions in Graphite Print Direct Evidence of Dirac Fermions in Graphite Print The recent surge of interest in the electronic properties of graphene-that is, isolated layers of graphite just one atomic layer thick-has largely been driven by the discovery that electron mobility in graphene is ten times higher than in commercial-grade silicon, raising the possibility of high-efficiency, low-power, carbon-based electronics. Scientists attribute graphene's surprising current capacity (as well as a number of even stranger phenomena) to the presence of charge carriers that behave as if they are massless, "relativistic" quasiparticles called Dirac fermions. Harnessing these quasiparticles in real-world carbon-based devices, however, requires a deeper knowledge of their behavior under less-than-ideal circumstances, such as around defects, at edges, or in three dimensions-in other words, in graphite. At the ALS, a team of researchers using angle-resolved photoemission spectroscopy (ARPES) have now produced the first direct evidence of massless Dirac fermions in graphite coexisting with quasiparticles of finite effective mass and defect-induced localized states.

336

Is Graphite a Diamonds Best Friend? New Information on Material  

NLE Websites -- All DOE Office Websites (Extended Search)

November 18th, 2003 November 18th, 2003 Is Graphite a Diamond's Best Friend? New Information on Material Transformation Science has yet to achieve the alchemist's dream of turning lead into gold. But a group of re-searchers using the GeoSoilEn-viroCARS (GSECARS) and High-Pressure Collaborative Access Team (HP-CAT) facilities at the Department of Energy's Advanced Photon Source (APS) at Argonne National Laboratory, may have found a way to turn ordinary soft graphite (source of the "lead" found in pencils) into a new, super-hard material that "looks" just like diamond. Using the high-brilliance x-ray beams from the APS, the group discovered that, under extreme pressure, graphite (among the softest of materials and the source of the lead found in pencils) becomes as hard as diamond, the

337

Composition and method for brazing graphite to graphite  

DOE Patents (OSTI)

The present invention is directed to a brazing material for joining graphite structures that can be used at temperatures up to about 2800.degree. C. The brazing material formed of a paste-like composition of hafnium carbide and uranium oxide with a thermosetting resin. The uranium oxide is converted to uranium dicarbide during the brazing operation and then the hafnium carbide and uranium dicarbide form a liquid phase at a temperature about 2600.degree. C. with the uranium diffusing and vaporizing from the joint area as the temperature is increased to about 2800.degree. C. so as to provide a brazed joint consisting essentially of hafnium carbide. This brazing temperature for hafnium carbide is considerably less than the eutectic temperature of hafnium carbide of about 3150.degree. C. The brazing composition also incorporates the thermosetting resin so that during the brazing operation the graphite structures may be temporarily bonded together by thermosetting the resin so that machining of the structures to final dimensions may be completed prior to the completion of the brazing operation. The resulting brazed joint is chemically and thermally compatible with the graphite structures joined thereby and also provides a joint of sufficient integrity so as to at least correspond with the strength and other properties of the graphite.

Taylor, Albert J. (Ten Mile, TN); Dykes, Norman L. (Oak Ridge, TN)

1984-01-01T23:59:59.000Z

338

Destruction of nuclear graphite using closed chamber incineration  

Science Conference Proceedings (OSTI)

Closed chamber incineration (CCI) is a novel technique by which irradiated nuclear graphite may be destroyed without the risk of radioactive cation release into the environment. The process utilizes an enclosed combustion chamber coupled with molten carbonate fuel cells (MCFCs). The transport of cations is intrinsically suppressed by the MCFCs, such that only the combustion gases are conducted through for release to the environment. An example CCI design was developed which had as its goal the destruction of graphite fuel elements from the Fort St. Vrain reactor (FSVR). By employing CCI, the volume of high level waste from the FSVR will be reduced by approximately 87 percent. Additionally, the incineration process will convert the SiC coating on the FSVR fuel particles to SiO{sub 2}, thus creating a form potentially suitable for direct incorporation in a vitrification process stream. The design is compact, efficient, and makes use of currently available technology.

Senor, D.J.; Hollenberg, G.W.; Morgan, W.C. [Pacific Northwest Lab., Richland, WA (United States); Marianowski, L.G. [Institute of Gas Technology, Chicago, IL (United States)

1994-03-01T23:59:59.000Z

339

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

Science Conference Proceedings (OSTI)

Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOE's structural materials research activities being conducted to support VHTR development. By far, the largest portion of material's R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water rea

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

2008-08-01T23:59:59.000Z

340

FY-09 Report: Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents  

Science Conference Proceedings (OSTI)

The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is performing research and development that focuses on key phenomena important during potential scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Gen-IV very high temperature reactor (VHTR). Phenomena Identification and Ranking Studies to date have identified that an air ingress event following on the heels of a VHTR depressurization is a very important incident. Consequently, the development of advanced air ingress-related models and verification and validation data are a very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air will enter the core through the break, leading to oxidation of the in-core graphite structure and fuel. If this accident occurs, the oxidation will accelerate heat-up of the bottom reflector and the reactor core and will eventually cause the release of fission products. The potential collapse of the core bottom structures causing the release of CO and fission products is one of the concerns. Therefore, experimental validation with the analytical model and computational fluid dynamic (CFD) model developed in this study is very important. Estimating the proper safety margin will require experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model. It will also require effective strategies to mitigate the effects of oxidation. The results from this research will provide crucial inputs to the INL NGNP/VHTR Methods Research and Development project. The second year of this three-year project (FY-08 to FY-10) was focused on (a) the analytical, CFD, and experimental study of air ingress caused by density-driven, stratified, countercurrent flow; (b) advanced graphite oxidation experiments and modeling; (c) experimental study of burn-off in the core bottom structures, (d) implementation of advanced graphite oxidation models into the GAMMA code, and (f) air ingress and oxidation mitigation analyses of the whole air-ingress scenario.

Chang H. Oh; Eung S. Kim

2009-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Theoretical analysis of the subcritical experiments performed in the IPEN/MB-01 research reactor facility  

SciTech Connect

The theoretical analysis of the subcritical experiments performed at the IPEN/MB-01 reactor employing the coupled NJOY/AMPX-II/TORT systems was successfully accomplished. All the analysis was performed employing ENDF/B-VII.0. The theoretical approach follows all the steps of the subcritical model of Gandini and Salvatores. The theory/experiment comparison reveals that the calculated subcritical reactivity is in a very good agreement to the experimental values. The subcritical index ({xi}) shows some discrepancies although in this particular case some work still have to be made to model in a better way the neutron source present in the experiments. (authors)

Lee, S. M.; Dos Santos, A. [Inst. de Pesquisas Energeticas e Nucleares, Cidade Universitaria, Av. Lineu Prestes, 2242, 05508-000 Sao Paulo - SP (Brazil)

2012-07-01T23:59:59.000Z

342

AIR COOLED NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

Fermi, E.; Szilard, L.

1958-05-27T23:59:59.000Z

343

Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters  

Science Conference Proceedings (OSTI)

Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies.

Konzek, G.J.

1983-07-01T23:59:59.000Z

344

A one-group parametric sensitivity analysis for the graphite isotope ratio method and other related techniques using ORIGEN 2.2  

E-Print Network (OSTI)

Several methods have been developed previously for estimating cumulative energy production and plutonium production from graphite-moderated reactors. The Graphite Isotope Ratio Method (GIRM) is one well-known technique. This method is based on the measurement of trace isotopes in the reactors graphite matrix to determine the change in their isotopic ratios due to burnup. These measurements are then coupled with reactor calculations to determine the total plutonium and energy production of the reactor. To facilitate sensitivity analysis of these methods, a one-group cross section and fission product yield library for the fuel and graphite activation products has been developed for MAGNOX-style reactors. This library is intended for use in the ORIGEN computer code, which calculates the buildup, decay, and processing of radioactive materials. The library was developed using a fuel cell model in Monteburns. This model consisted of a single fuel rod including natural uranium metal fuel, magnesium cladding, carbon dioxide coolant, and Grade A United Kingdom (UK) graphite. Using this library a complete sensitivity analysis can be performed for GIRM and other techniques. The sensitivity analysis conducted in this study assessed various input parameters including 235U and 238U cross section values, aluminum alloy concentration in the fuel, and initial concentrations of trace elements in the graphite moderator. The results of the analysis yield insight into the GIRM method and the isotopic ratios the method uses as well as the level of uncertainty that may be found in the system results.

Chesson, Kristin Elaine

2007-08-01T23:59:59.000Z

345

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel  

SciTech Connect

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed.

Deen, J.R.; Snelgrove, J.L. (Argonne National Lab., IL (United States)); Papastergiou, C. (National Center for Scientific Research, Athens (Greece))

1992-01-01T23:59:59.000Z

346

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel  

SciTech Connect

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed.

Deen, J.R.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Papastergiou, C. [National Center for Scientific Research, Athens (Greece)

1992-12-31T23:59:59.000Z

347

Composition and method for brazing graphite to graphite  

DOE Patents (OSTI)

A brazing material is described for joining graphite structures that can be used up to 2800/sup 0/C. The brazing material is formed of a paste-like composition of hafnium carbide and uranium oxide with a thermosetting resin. The uranium oxide is converted to uranium dicarbide during the brazing operation and then the hafnium carbide and uranium dicarbide form a liquid phase at a temperature about 2600/sup 0/C with the uranium diffusing and vaporizing from the joint area as the temperature is increased to about 2800/sup 0/C so as to provide a brazed joint consisting essentially of hafnium carbide. The resulting brazed joint is chemically and thermally compatible with the graphite structures.

Taylor, A.J.; Dykes, N.L.

1982-08-10T23:59:59.000Z

348

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

349

THERMAL NUCLEAR REACTOR  

DOE Patents (OSTI)

Nuclear reactors of the graphite moderated air cooled type in which canned slugs or rods of fissile material are employed are discussed. Such a reactor may be provided with a means for detecting dust particles in the exhausted air. The means employed are lengths of dust absorbent cord suspended in vertical holes in the shielding structure above each vertical coolant flow channel to hang in the path of the cooling air issuing from the channels, and associated spindles and drive motors for hauling the cords past detectors, such as Geiger counters, for inspecting the cords periodically. This design also enables detecting the individual channel in which a fault condition may have occurred.

Fenning, F.W.; Jackson, R.F.

1957-09-24T23:59:59.000Z

350

A New Class of Risk-Importance Measures to Support Reactor Aging Management and the Prioritization of Materials Degradation Research  

Science Conference Proceedings (OSTI)

As the US fleet of light water reactors ages, the risks of operation might be expected to increase. Although probabilistic risk assessment has proven a critical resource in risk-informed regulatory decision-making, limitations in current methods and models have constrained their prospective value in reactor aging management. These limitations stem principally from the use of static component failure rate models (which do not allow the impact of component aging on failure rates to be represented) and a very limited treatment of passive components (which would be expected to have an increasingly significant risk contribution in an aging system). Yet, a PRA captures a substantial knowledge base that could be of significant value in addressing plant aging. In this paper we will describe a methodology and a new class of risk importance measures that allow the use of an existing PRA model to support the management of plant aging, the prioritization of improvements to non-destructive examination and monitoring techniques, and the establishment of research emphases in materials science. This methodology makes use of data resources generated under the USNRC Proactive Management of Materials Degradation program which addresses the anticipated effects of numerous aging degradation mechanisms on a wide variety of component types.

Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y.

2010-06-07T23:59:59.000Z

351

Nuclear reactor shield including magnesium oxide  

DOE Patents (OSTI)

An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

Rouse, Carl A. (Del Mar, CA); Simnad, Massoud T. (La Jolla, CA)

1981-01-01T23:59:59.000Z

352

Design of a low enrichment, enhanced fast flux core for the Massachusetts Institute of Technology Research Reactor  

E-Print Network (OSTI)

Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating ...

Ellis, Tyler Shawn

2009-01-01T23:59:59.000Z

353

Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements  

Science Conference Proceedings (OSTI)

The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

J. D. Bess; T. L. Maddock; M. A. Marshall

2011-09-01T23:59:59.000Z

354

Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs  

SciTech Connect

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

1994-04-01T23:59:59.000Z

355

Quarterly technical report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, April--June 1975  

SciTech Connect

The current water reactor safety activities of ANC are accomplished in four programs. The Semiscale Program consists of small-scale nonnuclear thermal- hydraulic experiments for the generation of experimental data that can be applied to analytical models describing loss-of-coolant accident (LOCA) phenomena in water-cooled nuclear power plants. Emphasis is placed on acquiring system effects data from integral tests that characterize the most significant thermal- hydraulic phenomena during the depressurization (blowdown) and emergency cooling phase of a LOCA. The LOFT Program provides test data to support: (a) assessment and improvement of the analytical methods utilized for predicting the behavior of pressurized water reactors (PWR) under LOCA conditions; (b) evaluation of the performance of PWR engineered safety features (ESF), particularly the emergency core cooling system (ECCS); and (c) assessment of the quantitative margins of safety inherent in the performance of these safety features. The Thermal Fuels Behavior Program is a program designed to provide information on the behavior of reactor fuels under normal, off-normal, and accident conditions. The experimental portion is concentrated on testing of single fuel rods and fuel rod clusters under power-cooling-mismatch (PCM), loss-of-coolant, and reactivity initiated accident conditions. The Reactor Behavior Program encompasses the analytical aspects of predicting the response of nuclear power reactors under normal, abnormal, and accident conditions. The status of each program is reported. (auth)

1975-11-01T23:59:59.000Z

356

Microstructural Characterization of Nuclear Graphite  

Science Conference Proceedings (OSTI)

Symposium, Materials Solutions for the Nuclear Renaissance. Presentation Title ... Titanium Beryllide (TiBe12) for Fusion Reactors TRISO-coated Fuel...

357

Toughened Graphite Electrodes for Electric Arc Furnaces  

Benefits of Fiber Toughened Electrode Summary: Technology Description A method to more uniformly distribute graphite/carbon fibers into the electrode matrix by ...

358

AN EVALUATION OF THE URANIUM CONTAMINATION ON THE SURFACES OF ALCLAD URANIUM-ALUMINUM ALLOY RESEARCH REACTOR FUEL PLATES  

SciTech Connect

Reported radioactivity in the Low-Intensity Test Reactor (LITR) water coolant traceable to uranium contamination on the surfaces of the alclad uranium-- aluminum plate-tyne fuel element led to an investigation to determine the sources of uranium contamination on the fuel plate surfaces. Two possible contributors to surface contamination are external sources such as rolling-mill equipment, the most obvious, and diffusion of uranium from the uranium-aluminum alloy fuel into the aluminum cladding. This diffusion is likely because of the 600 deg C heat treatments used in the conventional fabrication process. Uranium determinations based on neutron activation analysis of machined layers from fuel plate surfaces showed that rolling-mill equipment, contaminated with highly enriched uranium, was responsible for transferring as much as 180 ppm U to plate surfaces. By careful practice where cleanliness is emphasized, surface contamination can be reduced to 0.6 ppm U/sup 235/. The residue remaining on the plate surface may be accounted for by diffusion of uranium from the fuel alloy into and through the cladding of the fuel plate. Data obtained from preliminary diffusion studies permitted a good estimate to be made of the diffusion coefficient of uranium into aluminum at 600 deg C: 2.5 x 10/sup -8/ cm//sec. To minimize diffusion while the plate-type aluminum-base research reactor fuel element is being processed, heat treatments at 600 deg C should be limited to 2.5 hr. The uranium contamination on the surfaces of the finished fuel plates should then be less than 0.6 ppm U / sup 235/ . This investigation also revealed that the solubility limit of uranium in aluminum at 600 deg C is approx 60 ppm. (auth)

Beaver, R.J.; Erwin, J.H.; Mateer, R.S.

1962-03-19T23:59:59.000Z

359

Applications of Nd:YAG laser micromanufacturing in High Temperature Gas Reactor research  

SciTech Connect

Two innovative applications of Nd:YAG laser micromachining techniques are demonstrated in this publication. Research projects to determine the fission product transport mechanisms in TRISO coated particles necessitate heat treatment studies as well as the manufacturing of a unique sealed system for experimentation at very high temperatures. This article describes firstly the design and creation of an alumina jig designed to contain 500 {mu}m diameter ZrO2 spheres intended for annealing experiments at temperatures up to 1600 C. Functional requirements of this jig are the precision positioning of spheres for laser ablation, welding and post weld heat treatment in order to ensure process repeatability and accurate indexing of individual spheres. The design challenges and the performance of the holding device are reported. Secondly the manufacture of a sealing system using laser micromachining is reported. ZrO2 micro plugs isolate the openings of micro-machined cavities to produce a gas-tight seal fit for application in a high temperature environment. The technique is described along with a discussion of the problems experienced during the sealing process. Typical problems experienced were seating dimensions and the relative small size ({approx} 200 {mu}m) of these plugs that posed handling challenges. Manufacturing processes for both the tapered seating cavity and the plug are demonstrated. In conclusion, this article demonstrates the application of Nd-YAG micromachining in an innovative way to solve practical research problems.

I. J. van Rooyen; C. A. Smal; J. Steyn; H. Greyling

2012-08-01T23:59:59.000Z

360

Effect of Vinylene Carbonate on Graphite Anode Cycling Efficiency  

E-Print Network (OSTI)

shift" (8). HydroQuebec's SNG-12 anode graphite was chosenElectrode laminates with SNG-12 graphite as the activePrevious experiments with SNG-12 graphite in coin cells with

Ridgway, Paul

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Graphitized-carbon fiber/carbon char fuel  

DOE Patents (OSTI)

A method for recovery of intact graphitic fibers from fiber/polymer composites is described. The method comprises first pyrolyzing the graphite fiber/polymer composite mixture and then separating the graphite fibers by molten salt electrochemical oxidation.

Cooper, John F. (Oakland, CA)

2007-08-28T23:59:59.000Z

362

STUDY OF REMOTE MILITARY POWER APPLICATIONS. REPORT NO. 12. EVALUATION AND SELECTION OF APPLICABLE REACTOR CONCEPTS  

SciTech Connect

An evaluation of the reactor concepts under consideration for remote military power plants is presented. The concepts include water-cooled and - moderated reactors, both direct and indirect cycle. organic-cooled and -moderated reactors, heavy-water-cooled and -moderated reactors. gas-cooled reactors, sodium- cooled graphite-moderated reactors, fast reactors, and fluid-fuel reactors. The limitations and advantages, technological status, economics, and future potential of each reactor are reviewed. From the reviews it is concluded that direct-cycle boiling-water and pressurized-water reactors are most suitable for applications requiring power before 1965. (C.J.G.)

1960-01-01T23:59:59.000Z

363

FINAL REPORT on Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy is performing research and development that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Generation IV very high temperature reactor (VHTR). Phenomena Identification and Ranking studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important. Consequently, the development of advanced air ingress-related models and verification & validation are of very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air ingress will occur through the break, leading to oxidation of the in-core graphite structure and fuel. This study indicates that depending on the location and the size of the pipe break, the air ingress phenomena are different. In an effort to estimate the proper safety margin, experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model are required. It will also require effective strategies to mitigate the effects of oxidation, eventually. This 3-year project (FY 2008FY 2010) is focused on various issues related to the VHTR air-ingress accident, including (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the core bottom structures, (d) structural tests of the oxidized core bottom structures, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i) verification and validation of the coupled models.

Chang H. Oh; Eung S. Kim; Hee C. NO; Nam Z. Cho

2011-01-01T23:59:59.000Z

364

GAS COOLED NUCLEAR REACTORS  

DOE Patents (OSTI)

A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

Long, E.; Rodwell, W.

1958-06-10T23:59:59.000Z

365

PIA - 10th International Nuclear Graphite Specialists Meeting...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

10th International Nuclear Graphite Specialists Meeting registration web site PIA - 10th International Nuclear Graphite Specialists Meeting registration web site PIA - 10th...

366

Graphite/Copper Composites from Natural Wood Precursors  

Science Conference Proceedings (OSTI)

Graphite derived from natural wood precursors provides a uniquely anisotropic porous scaffold for the fabrication of graphite/copper composites. The wettability...

367

Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation  

SciTech Connect

On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Re, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

2008-07-01T23:59:59.000Z

368

Seismic Safety Margins Research Programs. Assessment of potential increases in risk due to degradation of steam generator and reactor coolant pump supports. [PWR  

Science Conference Proceedings (OSTI)

During the NRC licensing review for the North Anna Units 1 and 2 pressurized-water reactors (PWRs), questions were raised regarding the potential for low-fracture toughness of steam-generator and reactor-coolant-pump supports. Because other PWRs may face similar problems, this issue was incorporated into the NRC Program for Resolution of Generic Issues. The work described in this report was performed to provide the NRC with a quantitative evaluation of the value/impact implications of the various options of resolving the fracture-toughness question. This report presents an assessment of the probabilistic risk associated with nil-ductility failures of steam-generator and reactor-coolant-pump structural-support systems during seismic events, performed using the Seismic Safety Margins Research Program codes and data bases.

Bohn, M. P.; Wells, J. E.; Shieh, L. C.; Cover, L. E.; Streit, R. L.

1983-08-01T23:59:59.000Z

369

Microstructural Characterization of Next Generation Nuclear Graphites  

Science Conference Proceedings (OSTI)

This article reports the microstructural characteristics of various petroleum and pitch based nuclear graphites (IG-110, NBG-18, and PCEA) that are of interest to the next generation nuclear plant program. Bright-field transmission electron microscopy imaging was used to identify and understand the different features constituting the microstructure of nuclear graphite such as the filler particles, microcracks, binder phase, rosette-shaped quinoline insoluble (QI) particles, chaotic structures, and turbostratic graphite phase. The dimensions of microcracks were found to vary from a few nanometers to tens of microns. Furthermore, the microcracks were found to be filled with amorphous carbon of unknown origin. The pitch coke based graphite (NBG-18) was found to contain higher concentration of binder phase constituting QI particles as well as chaotic structures. The turbostratic graphite, present in all of the grades, was identified through their elliptical diffraction patterns. The difference in the microstructure has been analyzed in view of their processing conditions.

Karthik Chinnathambi; Joshua Kane; Darryl P. Butt; William E. Windes; Rick Ubic

2012-04-01T23:59:59.000Z

370

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT  

SciTech Connect

Progress on reactor programs and in general engineering research and development programs is summarized. Research and development are reported on water-cooled reactors including EBWR and Borax-V, sodium-cooled reactors including ZPR-III, IV, and IX, Juggernaut, and EBR-I and II. Other work included a review of fast reactor technology, and studies on nuclear superheat, thermal and fast reactor safety, and reactor physics. Effort was also devoted to reactor materials and fuels development, heat engineering, separation processes and advanced reactor concepts. (J.R.D.)

1961-04-01T23:59:59.000Z

371

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT, MAY 1962  

SciTech Connect

Research progress is reported on water-cooled reactors, liquid-metal- cooled reactors, general reactor technology, plutonium recycle, advanced systems research and development, and nuclear safety. (M.C.G.)

1962-06-15T23:59:59.000Z

372

Retention of hydrogen in graphite  

DOE Green Energy (OSTI)

The retention of hydrogen in POCO AXF-5Q graphite has been measured at room temperature as a function of fluence and flux for H/sub 2//sup +/ ions at energies from 250 to 500 eV provided by a glow discharge. More than 2 x 10/sup 18/ H/cm/sup 2/ has been retained, and no indication of saturation has been observed to a fluence of 5 x 10/sup 19/ H/cm/sup 2/. In this experiment, retention was found to increase linearly with fluence for constant flux. A flux dependence was observed; that is, the retention rate was observed to decrease monotonically as the flux increased. A change-over experiment, deuterium to hydrogen, was conducted; the results show that significant change-over occurs (i.e., about 30% change-over for a fluence of 5 x 10/sup 17/ D/cm/sup 2/).

Langley, R.A.

1986-10-01T23:59:59.000Z

373

Graphite Oxidation Thermodynamics/Reactions  

Science Conference Proceedings (OSTI)

The vulnerability of graphite-matrix spent nuclear fuel to oxidation by the ambient atmosphere if the fuel canister is breached was evaluated. Thermochemical and kinetic data over the anticipated range of storage temperatures (200 to 400 C) were used to calculate the times required for a total carbon mass loss of 1 mgcm-2 from a fuel specimen. At 200 C, the time required to produce even this small loss is large, 900,000 yr. However, at 400 C the time required is only 1.9 yr. The rate of oxidation at 200 C is negligible, and the rate even at 400 C is so small as to be of no practical consequence. Therefore, oxidation of the spent nuclear fuel upon a loss of canister integrity is not anticipated to be a concern based upon the results of this study.

Propp, W.A.

1998-09-01T23:59:59.000Z

374

Method for molding threads in graphite panels  

DOE Patents (OSTI)

A graphite panel (10) with a hole (11) having a damaged thread (12) is repaired by drilling the hole (11) to remove all of the thread and make a new hole (13) of larger diameter. A bolt (14) with a lubricated thread (17) is placed in the new hole (13) and the hole (13) is packed with graphite cement (16) to fill the hole and the thread on the bolt. The graphite cement (16) is cured, and the bolt is unscrewed therefrom to leave a thread (20) in the cement (16) which is at least as strong as that of the original thread (12).

Short, William W. (Livermore, CA); Spencer, Cecil (Silverton, OR)

1994-01-01T23:59:59.000Z

375

Method for molding threads in graphite panels  

DOE Patents (OSTI)

A graphite panel with a hole having a damaged thread is repaired by drilling the hole to remove all of the thread and making a new hole of larger diameter. A bolt with a lubricated thread is placed in the new hole and the hole is packed with graphite cement to fill the hole and the thread on the bolt. The graphite cement is cured, and the bolt is unscrewed therefrom to leave a thread in the cement which is at least as strong as that of the original thread. 8 figures.

Short, W.W.; Spencer, C.

1994-11-29T23:59:59.000Z

376

Monte Carlo methods of neutron beam design for neutron capture therapy at the MIT Research Reactor (MITR-II)  

Science Conference Proceedings (OSTI)

Monte Carlo methods of coupled neutron/photon transport are being used in the design of filtered beams for Neutron Capture Therapy (NCT). This method of beam analysis provides segregation of each individual dose component, and thereby facilitates beam optimization. The Monte Carlo method is discussed in some detail in relation to NCT epithermal beam design. Ideal neutron beams (i.e., plane-wave monoenergetic neutron beams with no primary gamma-ray contamination) have been modeled both for comparison and to establish target conditions for a practical NCT epithermal beam design. Detailed models of the 5 MWt Massachusetts Institute of Technology Research Reactor (MITR-II) together with a polyethylene head phantom have been used to characterize approximately 100 beam filter and moderator configurations. Using the Monte Carlo methodology of beam design and benchmarking/calibrating our computations with measurements, has resulted in an epithermal beam design which is useful for therapy of deep-seated brain tumors. This beam is predicted to be capable of delivering a dose of 2000 RBE-cGy (cJ/kg) to a therapeutic advantage depth of 5.7 cm in polyethylene assuming 30 micrograms/g 10B in tumor with a ten-to-one tumor-to-blood ratio, and a beam diameter of 18.4 cm. The advantage ratio (AR) is predicted to be 2.2 with a total irradiation time of approximately 80 minutes. Further optimization work on the MITR-II epithermal beams is expected to improve the available beams. 20 references.

Clement, S.D.; Choi, J.R.; Zamenhof, R.G.; Yanch, J.C.; Harling, O.K. (Massachusetts Institute of Technology, Cambridge (USA))

1990-01-01T23:59:59.000Z

377

Diffusion of uranium in H-451 graphite at 900 to 1400/sup 0/C  

SciTech Connect

In this study, the diffusion of uranium (as a stand-in for plutonium) was investigated under conditions approximating those of the primary coolant loop in a High Temperature Gas-Cooled Reactor (HTGR). Profiles were obtained for uranium penetration in H-451 graphite (from the Great Lakes Carbon Company) at temperatures ranging from 900 to 1400/sup 0/C. Diffusion coefficients are established for UO/sub 2/ and UC/sub 2/.

Tallent, O.K.; Wichner, R.P.; Towns, R.L.

1983-03-01T23:59:59.000Z

378

On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor  

Science Conference Proceedings (OSTI)

IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

Ayman I. Hawari; Mohamed A. Bourham

2010-04-22T23:59:59.000Z

379

Developing fuel management capabilities based on coupled Monte Carlo depletion in support of the MIT Research Reactor (MITR) conversion  

E-Print Network (OSTI)

Pursuant to a 1986 NRC ruling, the MIT Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. Prior studies have shown that the MITR will be able to ...

Romano, Paul K. (Paul Kollath)

2009-01-01T23:59:59.000Z

380

Electrolytic Infiltration into Laser Sintered Porous Graphite  

Science Conference Proceedings (OSTI)

Symposium, Green Technologies for Materials Manufacturing and Processing V. Presentation Title, Electrolytic Infiltration into Laser Sintered Porous Graphite ... Tensile and Fatigue Testing of 304 Stainless Steel after Gaseous Hydrogen...

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Immobilization of Rocky Flats Graphite Fines Residue  

Science Conference Proceedings (OSTI)

The development of the immobilization process for graphite fines has proceeded through a series of experimental programs. The experimental procedures and results from each series of experiments are discussed in this report.

Rudisill, T.S.

1999-04-06T23:59:59.000Z

382

GRAPHITE PRODUCTION UTILIZING URANYL NITRATE HEXAHYDRATE CATALYST  

DOE Patents (OSTI)

ABS>The graphitizing of a mixture composed of furfuryl alcohol binder and uranyl nitrate hexahydrate hardener and the subsequent curing, baking, and graphitizing with pressure being initially applied prior to curing are described. The pressure step may be carried out by extrusion, methyl cellulose being added to the mixture before the completion of extrusion. Uranium oxide may be added to the graphitizable mixture prior to the heating and pressure steps. The graphitizable mixture may consist of discrete layers of different compositions. (AEC)

Sheinberg, H.; Armstrong, J.R.; Schell, D.H.

1964-03-10T23:59:59.000Z

383

Adsorption potential of alkanes on graphite  

SciTech Connect

In the framework of the extended Hueckel theory, the short-range repulsive interaction of alkanes with graphite is determined with band structure calculations from the difference between the total energy of the system (adsorbate + graphite) and the energy of the separated species. This theoretical approach enables one to determine the coefficients of the repulsive exponential term in the atom-atom potential simplified expression. The adsorption potential of alkanes on graphite is obtained when the dispersion atom-atom potential, which takes into account the high anisotropic polarizability of graphite, is added to the repulsive term. The equilibrium distance of methane on graphite and its vibrational frequency perpendicular to the surface are in good agreement with the experimental ones measured at low temperatures by neutron scattering techniques. The van der Waals radii of carbon and hydrogen atoms are obtained from the equilibrium distance of the atom-atom potential simplified expression. They are compared with those used in the literature to establish the semiempirical potential expressions. The molecular statistical theory of adsorption based on the atom-atom potential function enables one to predict the second adsorbate/surface virial coefficient and the thermodynamic characteristics of adsorption, measured for methane, ethane, and propane on graphitized carbon black at zero surface coverage by static and gas chromatographic methods.

Vidal-Madjar, C.; Minot, C.

1987-07-16T23:59:59.000Z

384

Biosciences Division: Endurance Bioenergy Reactor(tm)  

NLE Websites -- All DOE Office Websites (Extended Search)

Endurance Bioenergy Reactor(tm) DOE Logo Search BIO ... Search Argonne Home > BIO home > Endurance Bioenergy Reactor(tm) BIO Home Page About BIO News Releases Research Publications...

385

RERTR program activities related to the development and application of new LEU fuels. [Reduced Enrichment Research and Test Reactor; low-enriched uranium  

SciTech Connect

The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the current 1.7 g U/cm/sup 3/ to the 7.0 g U/cm/sup 3/ which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years.

Travelli, A.

1983-01-01T23:59:59.000Z

386

Graphite matrix materials for nuclear waste isolation  

SciTech Connect

At low temperatures, graphites are chemically inert to all but the strongest oxidizing agents. The raw materials from which artificial graphites are produced are plentiful and inexpensive. Morover, the physical properties of artificial graphites can be varied over a very wide range by the choice of raw materials and manufacturing processes. Manufacturing processes are reviewed herein, with primary emphasis on those processes which might be used to produce a graphite matrix for the waste forms. The approach, recommended herein, involves the low-temperature compaction of a finely ground powder produced from graphitized petroleum coke. The resultant compacts should have fairly good strength, low permeability to both liquids and gases, and anisotropic physical properties. In particular, the anisotropy of the thermal expansion coefficients and the thermal conductivity should be advantageous for this application. With two possible exceptions, the graphite matrix appears to be superior to the metal alloy matrices which have been recommended in prior studies. The two possible exceptions are the requirements on strength and permeability; both requirements will be strongly influenced by the containment design, including the choice of materials and the waste form, of the multibarrier package. Various methods for increasing the strength, and for decreasing the permeability of the matrix, are reviewed and discussed in the sections in Incorporation of Other Materials and Elimination of Porosity. However, it would be premature to recommend a particular process until the overall multi-barrier design is better defined. It is recommended that increased emphasis be placed on further development of the low-temperature compacted graphite matrix concept.

Morgan, W.C.

1981-06-01T23:59:59.000Z

387

Method of producing exfoliated graphite, flexible graphite, and nano-scaled graphene platelets  

DOE Patents (OSTI)

The present invention provides a method of exfoliating a layered material (e.g., graphite and graphite oxide) to produce nano-scaled platelets having a thickness smaller than 100 nm, typically smaller than 10 nm. The method comprises (a) dispersing particles of graphite, graphite oxide, or a non-graphite laminar compound in a liquid medium containing therein a surfactant or dispersing agent to obtain a stable suspension or slurry; and (b) exposing the suspension or slurry to ultrasonic waves at an energy level for a sufficient length of time to produce separated nano-scaled platelets. The nano-scaled platelets are candidate reinforcement fillers for polymer nanocomposites. Nano-scaled graphene platelets are much lower-cost alternatives to carbon nano-tubes or carbon nano-fibers.

Zhamu, Aruna (Centerville, OH); Shi, Jinjun (Columbus, OH); Guo, Jiusheng (Centerville, OH); Jang, Bor Z. (Centerville, OH)

2010-11-02T23:59:59.000Z

388

Utilization of MIT research reactor by Boston-area universities. Final report, July 1, 1978-June 30, 1980  

SciTech Connect

The guest institutions which used the MITR reactor during 1978-1979 are listed. The participation during previous years since the program began at MIT is also indicated. A summary of the type and amount of participation by local institutions is given.

Clark, L. Jr.; Janghorbani, M.

1981-07-01T23:59:59.000Z

389

TURRET: A HIGH TEMPERATURE GAS-CYCLE REACTOR PROPOSAL  

SciTech Connect

A nitrogen-cooled graphite-moderated nuclear reactor experiment is proposed to drive a closed-cycle gas turbine power plant at 1300 deg F. The annular core of the reactor can be rotated inside the reflector to permit fuel loading and discharge while operating at full power. Small cylindrical fuel elements of graphite are solutionimpregnated with partially enriched uranium. The fuel is recycled by incineration of the elements, chemical fresh graphite tn a small batch process. The unclad, uncoated fuel should permit high burn-up and simple fuel processing, but allows fission product diffusion into the gas stream. While methods are proposed for the removal of these from the gas, the Song-term consequences on turbine operation are unknown. The compatibility of nitrogen gas with the fuel has been studied experimentally. The radial movement of fuel gives a reactor with a constant power profile and no excess reactivity. The temperature is regulated by the fuel charging rate. (auth)

Hammond, R.P.; Busey, H.M.; Chapman, K.R.; Durham, F.P.; Rogers, J.D.; Wykoff, W.R.

1958-01-23T23:59:59.000Z

390

Summary report on the HFED (High-Uranium-Loaded Fuel Element Development) miniplate irradiations for the RERTR (Reduced Enrichment Research and Test Reactor) Program  

SciTech Connect

An experiment to evaluate the irradiation characteristics of various candidate low-enriched, high-uranium content fuels for research and test reactors was performed for the US Department of Energy Reduced Enrichment Research and Test Reactor Program. The experiment included the irradiation of 244 miniature fuel plates (miniplates) in a core position in the Oak Ridge Research Reactor. The miniplates were aluminum-based, dispersion-type plates 114.3 mm long by 50.8 mm wide with overall plate thicknesses of 1.27 or 1.52 mm. Fuel core dimensions varied according to the overall plate thicknesses with a minimum clad thickness of 0.20 mm. Tested fuels included UAl/sub x/, UAl/sub 2/, U/sub 3/O/sub 8/, U/sub 3/SiAl, U/sub 3/Si, U/sub 3/Si/sub 1.5/, U/sub 3/Si/sub 2/, U/sub 3/SiCu, USi, U/sub 6/Fe, and U/sub 6/Mn/sub 1.3/ materials. Although most miniplates were made with low-enriched uranium (19.9%), some with medium-enriched uranium (40 to 45%), a few with high-enriched uranium (93%), and a few with depleted uranium (0.2 to 0.4%) were tested for comparison. These fuel materials were irradiated to burnups ranging from /approximately/27 to 98 at. % /sup 235/U depletion. Operation of the experiment, measurement of miniplate thickness as the irradiation progressed, ultimate shipment of the irradiated miniplates to various hot cells, and preliminary results are reported here. 18 refs., 12 figs., 7 tabs.

Senn, R.L.

1989-04-01T23:59:59.000Z

391

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

Davidson, J.K.

1963-11-19T23:59:59.000Z

392

The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial  

SciTech Connect

Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor will approach two million kilograms with an associated volume of 2,500 cubic meters. The materials were activated by 14 MeV neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize, and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project, we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project.

Keith Rule; Erik Perry; Jim Chrzanowski; Mike Viola; Ron Strykowsky

2003-09-15T23:59:59.000Z

393

Sandia National Laboratories: Research: Facilities: Annular Core...  

NLE Websites -- All DOE Office Websites (Extended Search)

Annular Core Research Reactor facility Nuclear science photo At the Annular Core Research Reactor (ACRR) facility, Sandia researchers can subject various test objects to a mixed...

394

Reactor operations informal report, October 1994  

Science Conference Proceedings (OSTI)

This monthly progress report is divided into two parts. Part one covers the Brookhaven Medical Research Reactor and part two covers the Brookhaven High Flux Beam Reactor. Information is given for each reactor covering the following areas: reactor operation; instrumentation; mechanical maintenance; occurrence reports; and reactor safety.

Hauptman, H.M.; Petro, J.N.; Jacobi, O.; Lettieri, V.; Holden, N.; Ports, D.; Petricek, R.

1994-10-01T23:59:59.000Z

395

Effect of changes in DOE pricing policies for enrichment and reprocessing on research reactor fuel cycle costs  

SciTech Connect

Fuel cycle costs with HEU and LEU fuels for the IAEA generic 10 MW reactor are updated to reflect the change in DOE pricing policy for enrichment services as of October 1985 and the published charges for LEU reprocessing services as of February 1986. The net effects are essentially no change in HEU fuel cycle costs and a reduction of about 8 to 10% in the fuel cycle costs for LEU silicide fuel.

Matos, J.E.; Freese, K.E.

1986-11-03T23:59:59.000Z

396

Research into the pyrolysis of pure cellulose, lignin, and birch wood flour in the China Lake entrained-flow reactor  

DOE Green Energy (OSTI)

This experimental program used the China Lake entrained-flow pyrolysis reactor to briefly investigate the pyrolysis of pure cellulose, pure lignin, and birch wood flour. The study determined that the cellulose and wood flour do pyrolyze to produce primarily gaseous products containing significant amounts of ethylene and other useful hydrocarbons. During attempts to pyrolyze powdered lignin, the material melted and bubbled to block the reactor entrance. The pure cellulose and wood flour produced C/sub 2/ + yields of 12% to 14% by weight, which were less than yields from an organic feedstock derived from processed municipal trash. The char yields were 0.1% by weight from cellulose and 1.5% from birch wood flour - one to two orders of magnitude less than were produced from the trash-derived feedstock. In scanning electron microscope photographs, most of the wood flour char had a sintered and agglomerated appearance, although some particles retained the gross cell characteristics of the wood flour. The appearance of the char particles indicated that the material had once been molten and possibly vapor before it formed spheroidal particles about 1 ..mu..m diameter which agglomerated to form larger char particles. The ability to completely melt or vaporize lignocellulosic materials under conditions of high heating rates has now been demonstrated in a continuous flow reactor and promises new techniques for fast pyrolysis. This char was unexpectedly attracted by a magnet, presumably because of iron contamination from the pyrolysis reactor tube wall. The production of water-insoluble tars was negligible compared to the tars produced from trash-derived feedstock. The production of water-soluble organic materials was fairly low and qualitatively appeared to vary inversely with temperature. This study was of a preliminary nature and additional studies are necessary to optimize ethylene production from these feedstocks.

Diebold, J.

1980-06-01T23:59:59.000Z

397

Graphite oxidation modeling for application in MELCOR.  

SciTech Connect

The Arrhenius parameters for graphite oxidation in air are reviewed and compared. One-dimensional models of graphite oxidation coupled with mass transfer of oxidant are presented in dimensionless form for rectangular and spherical geometries. A single dimensionless group is shown to encapsulate the coupled phenomena, and is used to determine the effective reaction rate when mass transfer can impede the oxidation process. For integer reaction order kinetics, analytical expressions are presented for the effective reaction rate. For noninteger reaction orders, a numerical solution is developed and compared to data for oxidation of a graphite sphere in air. Very good agreement is obtained with the data without any adjustable parameters. An analytical model for surface burn-off is also presented, and results from the model are within an order of magnitude of the measurements of burn-off in air and in steam.

Gelbard, Fred

2009-01-01T23:59:59.000Z

398

PNNL: Research  

NLE Websites -- All DOE Office Websites (Extended Search)

To achieve higher power levels, researchers now are turning their attention to advanced fuel designs. PNNL is developing a new metal fuel for light water reactors intended...

399

NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

Not Available

1994-08-01T23:59:59.000Z

400

NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

Not Available

1994-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "graphite research reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

The use of U/sub 3/Si/sub 2/ dispersed in aluminum in plate-type fuel elements for research and test reactors  

SciTech Connect

A high-density fuel based on U/sub 3/Si/sub 2/ dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U/sub 3/Si/sub 2/ fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U/sub 3/Si/sub 2/ particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U/sub 3/Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U/sub 3/Si/sub 2/-aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m/sup 3/ is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs.

Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

1987-10-01T23:59:59.000Z

402

Twenty-First Water Reactor Safety Information Meeting. Volume 3, Primary system integrity; Aging research, products and applications; Structural and seismic engineering; Seismology and geology: Proceedings  

SciTech Connect

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25-27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database.

Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)] [comp.; Brookhaven National Lab., Upton, NY (United States)

1994-04-01T23:59:59.000Z

403

Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments  

Science Conference Proceedings (OSTI)

A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy`s (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors.

Coles, G.A.; McKay, S.L.

1995-06-01T23:59:59.000Z

404

Method of making segmented pyrolytic graphite sputtering targets  

DOE Patents (OSTI)

Anisotropic pyrolytic graphite wafers are oriented and bonded together such that the graphite's high thermal conductivity planes are maximized along the back surface of the segmented pyrolytic graphite target to allow for optimum heat conduction away from the sputter target's sputtering surface and to allow for maximum energy transmission from the target's sputtering surface.

McKernan, Mark A. (Livermore, CA); Alford, Craig S. (Tracy, CA); Makowiecki, Daniel M. (Livermore, CA); Chen, Chih-Wen (Livermore, CA)

1994-01-01T23:59:59.000Z

405

HTGR Dust Safety Issues and Needs for Research and Development  

Science Conference Proceedings (OSTI)

This report presents a summary of high temperature gas-cooled reactor dust safety issues. It draws upon a literature review and the proceedings of the Very High Temperature Reactor Dust Assessment Meeting held in Rockville, MD in March 2011 to identify and prioritize the phenomena and issues that characterize the effect of carbonaceous dust on high temperature reactor safety. It reflects the work and input of approximately 40 participants from the U.S. Department of Energy and its National Labs, the U.S. Nuclear Regulatory Commission, industry, academia, and international nuclear research organizations on the topics of dust generation and characterization, transport, fission product interactions, and chemical reactions. The meeting was organized by the Idaho National Laboratory under the auspices of the Next Generation Nuclear Plant Project, with support from the U.S. Nuclear Regulatory Commission. Information gleaned from the report and related meetings will be used to enhance the fuel, graphite, and methods technical program plans that guide research and development under the Next Generation Nuclear Plant Project. Based on meeting discussions and presentations, major research and development needs include: generating adsorption isotherms for fission products that display an affinity for dust, investigating the formation and properties of carbonaceous crust on the inside of high temperature reactor coolant pipes, and confirming the predominant source of dust as abrasion between fuel spheres and the fuel handling system.

Paul W. Humrickhouse

2011-06-01T23:59:59.000Z

406

Research Article Continuous Production of Lipase-Catalyzed Biodiesel in a Packed-Bed Reactor: Optimization and Enzyme Reuse Study  

E-Print Network (OSTI)

Copyright 2011 Hsiao-Ching Chen et al. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited. An optimal continuous production of biodiesel by methanolysis of soybean oil in a packed-bed reactor was developed using immobilized lipase (Novozym 435) as a catalyst in a tert-butanol solvent system. Response surface methodology (RSM) and Box-Behnken design were employed to evaluate the effects of reaction temperature, flow rate, and substrate molar ratio on the molar conversion of biodiesel. The results showed that flow rate and temperature have significant effects on the percentage of molar conversion. On the basis of ridge max analysis, the optimum conditions were as follows: flow rate 0.1 mL/min, temperature 52.1 ? C, and substrate molar ratio 1: 4. The predicted and experimental values of molar conversion were 83.31 2.07 % and 82.81 .98%, respectively. Furthermore, the continuous process over 30 days showed no appreciable decrease in the molar conversion. The paper demonstrates the applicability of using immobilized lipase and a packed-bed reactor for continuous biodiesel synthesis. 1.

Hsiao-ching Chen; Hen-yi Ju; Tsung-ta Wu; Yung-chuan Liu; Chih-chen Lee; Cheng Chang; Yi-lin Chung; Chwen-jen Shieh

2010-01-01T23:59:59.000Z

407

The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases  

SciTech Connect

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest status and plans are presented.

Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

2012-10-01T23:59:59.000Z

408

Carbon-14 Bioassay for Decommissioning of Hanford Reactors  

SciTech Connect

The old production reactors at the US Department of Energy Hanford Site used large graphite piles as the moderator. As part of long-term decommissioning plans, the potential need for 14C radiobioassay of workers was identified. Technical issues associated with 14C bioassay and worker monitoring were investigated, including anticipated graphite characterization, potential intake scenarios, and the bioassay capabilities that may be required to support the decommissioning of the graphite piles. A combination of urine and feces sampling would likely be required for the absorption type S 14C anticipated to be encountered. However the concentrations in the graphite piles appear to be sufficiently low that dosimetrically significant intakes of 14C are not credible, thus rendering moot the need for such bioassay.

Carbaugh, Eugene H.; Watson, David J.

2012-05-01T23:59:59.000Z

409

STUDY ON AIR INGRESS MITIGATION METHODS IN THE VERY HIGH TEMPERATURE GAS COOLED REACTOR (VHTR)  

SciTech Connect

An air-ingress accident followed by a pipe break is considered as a critical event for a very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break leading to oxidation of the in-core graphite structure. Thus, without mitigation features, this accident might lead to severe exothermic chemical reactions of graphite and oxygen. Under extreme circumstances, a loss of core structural integrity may occur along with excessive release of radiological inventory. Idaho National Laboratory under the auspices of the U.S. Department of Energy is performing research and development (R&D) that focuses on key phenomena important during challenging scenarios that may occur in the VHTR. Phenomena Identification and Ranking Table (PIRT) studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Oh et al. 2006, Schultz et al. 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) requirements are part of the experimental validation plan. This paper discusses about various air-ingress mitigation concepts applicable for the VHTRs. The study begins with identifying important factors (or phenomena) associated with the air-ingress accident by using a root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air-ingress mitigation ideas can be conceptually derived. The main concepts include (1) preventing structural degradation of graphite supporters; (2) preventing local stress concentration in the supporter; (3) preventing graphite oxidation; (4) preventing air ingress; (5) preventing density gradient driven flow; (4) preventing fluid density gradient; (5) preventing fluid temperature gradient; (6) preventing high temperature. Based on the basic concepts listed above, various air-ingress mitigation methods are proposed in this study. Among them, the following two mitigation ideas are extensively investigated using computational fluid dynamic codes (CFD): (1) helium injection in the lower plenum, and (2) reactor enclosure opened at the bottom. The main idea of the helium injection method is to replace air in the core and the lower plenum upper part by buoyancy force. This method reduces graphite oxidation damage in the severe locations of the reactor inside. To validate this method, CFD simulations are addressed here. A simple 2-D CFD model is developed based on the GT-MHR 600MWt design. The simulation results showed that the helium replace the air flow into the core and significantly reduce the air concentration in the core and bottom reflector potentially protecting oxidation damage. According to the simulation results, even small helium flow was sufficient to remove air in the core, mitigating the air-ingress successfully. The idea of the reactor enclosure with an opening at the bottom changes overall air-ingress mechanism from natural convection to molecular diffusion. This method can be applied to the current system by some design modification of the reactor cavity. To validate this concept, this study also uses CFD simulations based on the simplified 2-D geometry. The simulation results showed that the enclosure open at the bottom can successfully mitigate air-ingress into the reactor even after on-set natural circulation occurs.

Chang H. Oh

2011-03-01T23:59:59.000Z

410

THERMIONIC SPACE POWER REACTOR SYSTEM RESEARCH AND DEVELOPMENT. Annual Summary Report, May 1, 1962-April 30, 1963  

SciTech Connect

Activities in a program to test and develop prototype fission-heated thermionic cells for space uses are reported. During the period, in-reactor tests were conducted on two W-clad U-bearing fuel emitters and one unclad type. Fuel emitter proof tests were also conducted which demonstrated 1000-hr operational capability of W-clad systems. Output power density and the temperature of heat rejection were found to have major effects on the weight- performance characteristics of the system. Advances in techniques related to W vapor deposition are reported. Descriptions of the fuel-emitter development, cell design and development, and testing of out-of-pile and in-pile cells are included. Operation of the clad-type test cells at design power and temperature led to selection of these cells for planned long-duration in-pile tests. (J.R.D.)

Elsner, N.B.; Holland, J.W.; Pidd, R.W.; Ream, J.T. Jr.; Wright, W.B.; Yang, L.

1963-12-17T23:59:59.000Z

411

FUEL CYCLE PROGRAM, A BOILING WATER REACTOR RESEARCH DEVELOPMENT PROGRAM. First Summary Report for March 1959-July 1960  

SciTech Connect

The Fuel Cycle Development Program is a basic development program for boiling and other water technology. It covers the areas of oxide fuel fabrication. irradiation. and examination; the physics of water-moderated reactore; and boiling-water heat transfer and stability. Schedules for the fuel- cycle program were examined. and it was concluded that portions of the Task A program should be conducted during the period May to Dec. 1959 in order to keep costs of the work as low as possible and to allow initiation of the fuel-cycle program at the earliest possible date after the Vallecitos BWR was returned to service. The basis for the scheduling of the work is discussed. and a chronological summary describing the content of the work is given. Technical progress is outlined and details are summarized. Subsequent reports issued monthly and quarterly will summarize the progress of the prognam. (W.D.M.)

Cook, W.H.

1961-10-31T23:59:59.000Z

412

Analysis of Natural Graphite, Synthetic Graphite, and Thermosetting Resin Candidates for Use in Fuel Compact Matrix  

Science Conference Proceedings (OSTI)

The AGR-1 and AGR-2 compacting process involved overcoating TRISO particles and compacting them in a steel die. The overcoating step is the process of applying matrix to the OPyC layer of TRISO particles in a rotating drum in order to build up an overcoat layer of desired thickness. The matrix used in overcoating is a mixture of natural graphite, synthetic graphite, and thermosetting resin in the ratio, by weight, of 64:16:20. A wet mixing process was used for AGR-1 and AGR-2, in that the graphites and resin were mixed in the presence of ethyl alcohol. The goal of the wet mixing process was to 'resinate' the graphite particles, or coat each individual graphite particle with a thin layer of resin. This matrix production process was similar to the German, Chinese, Japanese, and South African methods, which also use various amount of solvent during mixing. See Appendix 1 for information on these countries matrix production techniques. The resin used for AGR-1 and AGR-2 was provided by Hexion, specifically Hexion grade Durite SC1008. Durite SC1008 is a solvated (liquid) resole phenolic resin. A resole resin does not typically have a hardening agent added. The major constituent of SC1008 is phenol, with minor amounts of formaldehyde. Durite SC1008 is high viscosity, so additional ethyl alcohol was added during matrix production in order to reduce its viscosity and enhance graphite particle resination. The current compacting scale up plan departs from a wet mixing process. The matrix production method specified in the scale up plan is a co-grinding jet mill process where powdered phenolic resin and graphite are all fed into a jet mill at the same time. Because of the change in matrix production style, SC1008 cannot be used in the jet milling process because it is a liquid. The jet milling/mixing process requires that a suite of solid or powdered resins be investigated. The synthetic graphite used in AGR-1 and AGR-2 was provided by SGL Carbon, grade KRB2000. KRB2000 is a graphitized petroleum coke. The availability of KRB2000 is perhaps in question, so a replacement synthetic graphite may need to be identified. This report presents data on potential replacements for KRB2000.

Trammell, Michael P [ORNL; Pappano, Peter J [ORNL

2011-09-01T23:59:59.000Z

413

Fuzzy control for nuclear reactor operation -- strengths, weaknesses, opportunities and threats  

Science Conference Proceedings (OSTI)

Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be analyzed through an analysis of strengths, ... Keywords: Artificial Intelligence, Br1 Reactor, Fuzzy Control, Reactor Operation, Triga Mark Iii Reactor

Da Ruan; Jorge S. Bentez-Read

2005-10-01T23:59:59.000Z

414

Generation -IV Reactor Concepts  

NLE Websites -- All DOE Office Websites (Extended Search)

Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),

415

Fast Breeder Reactors for Energy Security  

Science Conference Proceedings (OSTI)

Technical Paper / NSF Workshop on the Research Needs of the Next Generation Nuclear Power Technology / Fission Reactor

William M. Jacobi

416

A STUDY OF THE PRIMARY SHIELD FOR THE PRDC REACTOR  

SciTech Connect

Temperature distributions, irradiation effects, stacking arrangements, voidage, and economics for the borated-graphite shield of the PRDC reactor were investigated. Of the shield systems considered, four are reported here. System 1 contalns 30 in. of 1% borated graphite, with either ordinary graphite or a cement as a filler for the remaindcr of the volume. The maximum temperature at the flex plates in this system was calculated to be 5OO deg F. Systems 2 and 3 consist of 2 in. of 5% borated graphite near the core vessel and 1/2 in. of Boral at the primary-shield tank. A filler material of carbon blocks is used in System 2 and graphite in System 3. The calculated maximum temperatures were 700 deg F and 35O deg F, respectively. System 4 consists of a laminated structure of Boral and graphite near the primary-shield tank and carbon-block filler. It was calculated to have a maximum temperature of 600 deg F at the flex plates. The maximum temperature at the flex plates recommended by APDA is 500 deg F. Energy storage and radiation damage were found to be within permissible limits in all four systems. However, these conclusions are based on experimental data from the Hanford reactor in which the neutron-ene