10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2014 CFR

2014-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2012 CFR

2012-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2011 CFR

2011-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2013 CFR

2013-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

76 FR 14747 - Revision of Fee Schedules; Fee Recovery for Fiscal Year 2011

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-17

... Standards VI. Environmental Impact: Categorical Exclusion VII. Paperwork Reduction Act Statement VIII... nuclear reactors. The NRC has prepared a paper for the Commission's information in support of the Nuclear Energy Institute's position to calculate annual fees for each new licensed power reactor as a function of...

Decommissioning of eight surplus production reactors at the Hanford Site, Richland, Washington. Addendum (Final Environmental Impact Statement)

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

Not Available

1992-12-01

The first section of this volume summarizes the content of the draft environmental impact statement (DEIS) and this Addendum, which together constitute the final environmental impact statement (FEIS) prepared on the decommissioning of eight surplus plutonium production reactors at Hanford. The FEIS consists of two volumes. The first volume is the DEIS as written. The second volume (this Addendum) consists of a summary; Chapter 9, which contains comments on the DEIS and provides DOE`s responses to the comments; Appendix F, which provides additional health effects information; Appendix K, which contains costs of decommissioning in 1990 dollars; Appendix L, which containsmore » additional graphite leaching data; Appendix M, which contains a discussion of accident scenarios; Appendix N, which contains errata; and Appendix 0, which contains reproductions of the letters, transcripts, and exhibits that constitute the record for the public comment period.« less

Environmental Impact Statement. Space Nuclear Thermal Propulsion Program. Particle Bed reactor Propulsion Technology Development and Validation

DTIC Science & Technology

1993-05-01

further examination or disposal. 2.2.2.3 Non -Nucleer Engine Integration Tests. EITs would be designed to demonstrate proper function of the propellant...located 42 miles southwest of the CTF, is designated as a Class I air quality region. The nearest non -attainment area is Pocatello, Idaho, 75 miles south of...accelerate. combustiomn. Nintrogen and helnee are Staple CEnergy suggestsa design eail ofAt a em-le 500 aenIIirons 1 aoph,,,,ants ad non -reastive. ?annual

Farmers' beliefs about bovine tuberculosis control in Northern Ireland.

PubMed

O'Hagan, M J H; Matthews, D I; Laird, C; McDowell, S W J

2016-06-01

Beliefs can play an important role in farmer behaviour and willingness to adopt new policies. In Northern Ireland, bovine tuberculosis (bTB) is one of the most important endemic diseases facing the cattle industry. An observational study was conducted on 192 farms in a high bTB incidence area during 2010-2011 in order to obtain a better understanding of farmers' beliefs in relation to bTB control. The views of farmers who had experienced a recent confirmed or multiple reactor bTB breakdowns (cases) were compared to those of farmers who had no recent reactors or restricted herd tests (controls). Data were obtained from a face-to-face questionnaire assessing farmers' agreement to 22 statements. All participating farmers found bTB control important and most were keen to learn more about bTB biosecurity measures and were in favour of the cattle-related bTB control measures as presented in the questionnaire (isolation of skin test inconclusive animals, use of the gamma-interferon test and pre-movement testing). The majority of farmers would allow badger vaccination and culling on their own land with an overall preference for vaccination. Highest disagreement was shown for the statements querying a willingness to pay for bTB control measures. There was agreement on most issues between case and control farmers and between different age groups of farmers although case farmers showed more support for additional advice on bTB biosecurity measures (P = 0.042). Case farmers were also more in favour of allowing badger vaccination (P = 0.008) and culling (P = 0.043) on their land and showed less concern for public opposition (P = 0.048). Copyright © 2015 Elsevier Ltd. All rights reserved.

Experience of the nuclear reactors (environmental impact assessment for decommissioning) regulations 1999, as amended, in Great Britain

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

Brown, Sarah; Mattress, Elaine; Nettleton, Jo

2007-07-01

Available in abstract form only. Full text of publication follows: In Great Britain, the Nuclear Reactors (Environmental Impact Assessment for Decommissioning) Regulations 1999 as amended 2006 (EIADR) requires assessment of the potential environmental impacts of projects to decommission nuclear power stations and reactors. The Health and Safety Executive (HSE) is the competent authority for EIADR. The EIADR implement European Council Directive 85/337/EEC (the EIA Directive) as amended by Council Directive 97/11/EC and Council Directive 2003/35/EC the (Public Participation Directive). The purpose of the EIADR is to assess environmental effects of nuclear reactor decommissioning projects, involve the public through consultation, andmore » make the decision-making process open and transparent. Under the regulations, any licensee wishing to begin to decommission or dismantle a nuclear power station, or other civil nuclear reactor, must apply to HSE for consent to carry out the decommissioning project, undertake an environmental impact assessment and prepare an environmental statement that summarises the environmental effects of the project. HSE will consult on the environmental statement. So far under the EIADR there have been six consents granted for decommissioning projects for Magnox Power Stations. These stations have been required as a condition of consent to submit an Environmental Management Plan on an annual basis. This allows the project to be continually reviewed and assessed to ensure that the licensee can provide detail as agreed during the review of the environmental statement and that any changes to mitigation measures are detailed. This paper summarises the EIADR process, giving particular emphasis to public participation and the decision making process, and discusses HSE's experience of EIADR with reference to specific environmental issues raised by stakeholders and current developments. (authors)« less

Draft environmental impact statement siting, construction, and operation of New Production Reactor capacity. Volume 4, Appendices D-R

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

None

1991-04-01

This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build newmore » production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains 15 appendices.« less

Status of DOE efforts to renew acceptance of foreign research reactor spent nuclear fuel

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

Head, C.R.

1997-08-01

This presentation summarizes the efforts being made by the Department of Energy to renew acceptance of spent nuclear fuel shipments from foreign research reactors. The author reviews the actions undertaken in this process in a fairly chronological manner, through the present time, as well as the development of an environmental impact statement to support the proposed actions.

10 CFR 51.52 - Environmental effects of transportation of fuel and waste-Table S-4.

Code of Federal Regulations, 2011 CFR

2011-01-01

... thermal power level not exceeding 3,800 megawatts; (2) The reactor fuel is in the form of sintered uranium... Nuclear Power Plants,” WASH-1238, December 1972, and Supp. 1 NUREG-75/038 April 1975. Both documents are...-water-cooled nuclear power reactor, and submitted after February 4, 1975, shall contain a statement...

10 CFR 51.52 - Environmental effects of transportation of fuel and waste-Table S-4.

Code of Federal Regulations, 2014 CFR

2014-01-01

... thermal power level not exceeding 3,800 megawatts; (2) The reactor fuel is in the form of sintered uranium... Nuclear Power Plants,” WASH-1238, December 1972, and Supp. 1 NUREG-75/038 April 1975. Both documents are...-water-cooled nuclear power reactor, and submitted after February 4, 1975, shall contain a statement...

10 CFR 51.52 - Environmental effects of transportation of fuel and waste-Table S-4.

Code of Federal Regulations, 2012 CFR

2012-01-01

... thermal power level not exceeding 3,800 megawatts; (2) The reactor fuel is in the form of sintered uranium... Nuclear Power Plants,” WASH-1238, December 1972, and Supp. 1 NUREG-75/038 April 1975. Both documents are...-water-cooled nuclear power reactor, and submitted after February 4, 1975, shall contain a statement...

10 CFR 51.52 - Environmental effects of transportation of fuel and waste-Table S-4.

Code of Federal Regulations, 2013 CFR

2013-01-01

... thermal power level not exceeding 3,800 megawatts; (2) The reactor fuel is in the form of sintered uranium... Nuclear Power Plants,” WASH-1238, December 1972, and Supp. 1 NUREG-75/038 April 1975. Both documents are...-water-cooled nuclear power reactor, and submitted after February 4, 1975, shall contain a statement...

78 FR 66385 - Omaha Public Power District Fort Calhoun Station, Unit 1; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-05

... Nuclear Energy Institute (NEI) 06-11, ``Managing Personnel Fatigue at Nuclear Power Reactor Sites...), no environmental impact statement or environmental assessment is required to be prepared in...

A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility

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

S. Khericha

2010-12-01

The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn thesemore » actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3 to 5 years to construct.« less

Design data needs modular high-temperature gas-cooled reactor. Revision 2

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

NONE

1987-03-01

The Design Data Needs (DDNs) provide summary statements for program management, of the designer`s need for experimental data to confirm or validate assumptions made in the design. These assumptions were developed using the Integrated Approach and are tabulated in the Functional Analysis Report. These assumptions were also necessary in the analyses or trade studies (A/TS) to develop selections of hardware design or design requirements. Each DDN includes statements providing traceability to the function and the associated assumption that requires the need.

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

NASA Astrophysics Data System (ADS)

1993-01-01

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

75 FR 25303 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-07

... (previously DG-1203) ``Containment Structural Integrity Evaluation for Internal Pressure Loadings Above Design... the time allotted to present oral statements can be obtained from the website cited above or by...

PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

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

S. T. Khericha

2007-04-01

The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed tomore » achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.« less

On a distinctive feature of problems of calculating time-average characteristics of nuclear reactor optimal control sets

NASA Astrophysics Data System (ADS)

Trifonenkov, A. V.; Trifonenkov, V. P.

2017-01-01

This article deals with a feature of problems of calculating time-average characteristics of nuclear reactor optimal control sets. The operation of a nuclear reactor during threatened period is considered. The optimal control search problem is analysed. The xenon poisoning causes limitations on the variety of statements of the problem of calculating time-average characteristics of a set of optimal reactor power off controls. The level of xenon poisoning is limited. There is a problem of choosing an appropriate segment of the time axis to ensure that optimal control problem is consistent. Two procedures of estimation of the duration of this segment are considered. Two estimations as functions of the xenon limitation were plot. Boundaries of the interval of averaging are defined more precisely.

76 FR 44377 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on AP1000...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-25

... Design Control Document (DCD). The Subcommittee will hear presentations by and hold discussions with the... statements can be obtained from the website cited above or by contacting the identified DFO. Moreover, in...

Interim Safe Storage of Plutonium Production Reactors at the US DOE Hanford Site - 13438

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

Schilperoort, Daryl L.; Faulk, Darrin

2013-07-01

Nine plutonium production reactors located on DOE's Hanford Site are being placed into an Interim Safe Storage (ISS) period that extends to 2068. The Environmental Impact Statement (EIS) for ISS [1] was completed in 1993 and proposed a 75-year storage period that began when the EIS was finalized. Remote electronic monitoring of the temperature and water level alarms inside the safe storage enclosure (SSE) with visual inspection inside the SSE every 5 years are the only planned operational activities during this ISS period. At the end of the ISS period, the reactor cores will be removed intact and buried inmore » a landfill on the Hanford Site. The ISS period allows for radioactive decay of isotopes, primarily Co-60 and Cs-137, to reduce the dose exposure during disposal of the reactor cores. Six of the nine reactors have been placed into ISS by having an SSE constructed around the reactor core. (authors)« less

Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 5. Public comments and Nuclear Regulatory Commission responses

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

Not Available

1976-08-01

Copies of 69 letters are presented commenting on the Draft Generic Environmental Statement (GESMO) WASH-1327 and the NRC's responses to the comments received from Federal, State and local agencies; environmental and public interest groups, members of the academic and industrial communities, and individual citizens. An index to these letters indicating the number assigned to each letter, the author, and organization represented, is provided in the Table of Contents.

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

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

NONE

1995-08-01

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized intomore » six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.« less

78 FR 8195 - Biweekly Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-05

... of the bases for the contention and a concise statement of the alleged facts or expert opinion which..., ``Allowable Value for Primary Containment and Drywell Isolation Instrumentation,'' Function 3.c, ``Reactor Core Isolation Cooling (RCIC) Steam Supply Line Pressure--Low.'' This TS allowable value will be...

Final environmental impact statement. Management of commercially generated radioactive waste. Volume 3. Public comments hearing board report

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

Not Available

1980-10-01

This EIS analyzes the significant environmental impacts that could occur if various technologies for management and disposal of high-level and transuranic wastes from commercial nuclear power reactors were to be developed and implemented. This EIS will serve as the environmental input for the decision on which technology, or technologies, will be emphasized in further research and development activities in the commercial waste management program. The action proposed in this EIS is to (1) adopt a national strategy to develop mined geologic repositories for disposal of commercially generated high-level and transuranic radioactive waste (while continuing to examine subseabed and very deepmore » hole disposal as potential backup technologies) and (2) conduct a R and D program to develop such facilities and the necessary technology to ensure the safe long-term containment and isolation of these wastes. The Department has considered in this statement: development of conventionally mined deep geologic repositories for disposal of spent fuel from nuclear power reactors and/or radioactive fuel reprocessing wastes; balanced development of several alternative disposal methods; and no waste disposal action. This volume contains written public comments and hearing board responses and reports offered on the draft statement.« less

76 FR 80409 - Draft Supplement 2 to Final Environmental Statement Related to the Operation of Watts Bar Nuclear...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-23

... image files of the NRC's public documents. If you do not have access to ADAMS or if there are problems.... Stephen J. Campbell, Chief, Watts Bar Special Projects Branch, Division of Operating Reactor Licensing...

78 FR 59659 - Correction to the Final Environmental Impact Statement/Overseas Environmental Impact Statement...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-27

... Statement/Overseas Environmental Impact Statement for Hawaii-Southern California Training and Testing AGENCY... Final Environmental Impact Statement (EIS)/Overseas EIS (OEIS) for Hawaii- Southern California Training.... Navy Training and Testing Activities in the Hawaii- Southern California Training and Testing Study Area...

10 CFR 51.95 - Postconstruction environmental impact statements.

Code of Federal Regulations, 2011 CFR

2011-01-01

... the storage of spent fuel for the nuclear power plant within the scope of the generic determination in... a license to store spent fuel at a nuclear power reactor after expiration of the operating or... Section 51.95 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL PROTECTION REGULATIONS FOR...

10 CFR 51.95 - Postconstruction environmental impact statements.

Code of Federal Regulations, 2012 CFR

2012-01-01

... the storage of spent fuel for the nuclear power plant within the scope of the generic determination in... a license to store spent fuel at a nuclear power reactor after expiration of the operating or... Section 51.95 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL PROTECTION REGULATIONS FOR...

Methods and strategies for future reactor safety goals

NASA Astrophysics Data System (ADS)

Arndt, Steven Andrew

There have been significant discussions over the past few years by the United States Nuclear Regulatory Commission (NRC), the Advisory Committee on Reactor Safeguards (ACRS), and others as to the adequacy of the NRC safety goals for use with the next generation of nuclear power reactors to be built in the United States. The NRC, in its safety goals policy statement, has provided general qualitative safety goals and basic quantitative health objectives (QHOs) for nuclear reactors in the United States. Risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) have been used as surrogates for the QHOs. In its review of the new plant licensing policy the ACRS has looked at the safety goals, as has the NRC. A number of issues have been raised including what the Commission had in mind when it drafted the safety goals and QHOs, how risk from multiple reactors at a site should be combined for evaluation, how the combination of a new and old reactor at the same site should be evaluated, what the criteria for evaluating new reactors should be, and whether new reactors should be required to be safer than current generation reactors. As part of the development and application of the NRC safety goal policy statement the Commissioners laid out the expectations for the safety of a nuclear power plant but did not address the risk associated with current multi-unit sites, potential modular reactor sites, and hybrid sites that could contain current generation reactors, new passive reactors, and/or modular reactors. The NRC safety goals and the QHOs refer to a "nuclear power plant," but do not discuss whether a "plant" refers to only a single unit or all of the units on a site. There has been much discussion on this issue recently due to the development of modular reactors. Additionally, the risk of multiple reactor accidents on the same site has been largely ignored in the probabilistic risk assessments (PRAs) done to date, and in most risk-informed analyses and discussions. This dissertation examines potential approaches to updating the safety goals that include the establishment of new quantitative safety goal associated with the comparative risk of generating electricity by viable competing technologies and modifications of the goals to account for multi-plant reactor sites, and issues associated with the use of safety goals in both initial licensing and operational decision making. This research develops a new quantitative health objective that uses a comparable benefit risk metric based on the life-cycle risk of the construction, operation and decommissioning of a comparable non-nuclear electric generation facility, as well as the risks associated with mining and transportation. This dissertation also evaluates the effects of using various methods for aggregating site risk as a safety metric, as opposed to using single plant safety goals. Additionally, a number of important assumptions inherent in the current safety goals, including the effect of other potential negative societal effects such as the generation of greenhouse gases (e.g., carbon dioxide) have on the risk of electric power production and their effects on the setting of safety goals, is explored. Finally, the role risk perception should play in establishing safety goals has been explored. To complete this evaluation, a new method to analytically compare alternative technologies of generating electricity was developed, including development of a new way to evaluate risk perception, and a new method was developed for evaluating the risk at multiple units on a single site. To test these modifications to the safety goals a number of possible reactor designs and configurations were evaluated using these new proposed safety goals to determine the goals' usefulness and utility. The results of the analysis showed that the modifications provide measures that more closely evaluate the potential risk to the public from the operation of nuclear power plants than the current safety goals, while still providing a straight-forward process for assessment of reactor design and operation.

From Confrontation to Cooperation: 8th International Seminar on Nuclear War

NASA Astrophysics Data System (ADS)

Zichichi, A.; Dardo, M.

1992-09-01

The Table of Contents for the full book PDF is as follows: * OPENING SESSION * A. Zichichi: Opening Statements * R. Nicolosi: Opening Statements * MESSAGES * CONTRIBUTIONS * "The Contribution of the Erice Seminars in East-West-North-South Scientific Relations" * 1. LASER TECHNOLOGY * "Progress in laser technology" * "Progress in laboratory high gain ICF: prospects for the future" * "Applications of laser in metallurgy" * "Laser tissue interactions in medicine and surgery" * "Laser fusion" * "Compact X-ray lasers in the laboratory" * "Alternative method for inertial confinement" * "Laser technology in China" * 2. NUCLEAR AND CHEMICAL SAFETY * "Reactor safety and reactor design" * "Thereotical analysis and numerical modelling of heat transfer and fuel migration in underlying soils and constructive elements of nuclear plants during an accident release from the core" * "How really to attain reactor safely" * "The problem of chemical weapons" * "Long terms genetic effects of nuclear and chemical accidents" * "Features of the brain which are of importance in understanding the mode of operation of toxic substances and of radiation" * "CO2 and ultra safe reactors" * 3. USE OF MISSILES * "How to convert INF technology for peaceful scientific purposes" * "Beating words into plowshares: a proposal for the peaceful uses of retired nuclear warheads" * "Some thoughts on the peaceful use of retired nuclear warheads" * "Status of the HEFEST project" * 4. OZONE * "Status of the ozone layer problem" * 5. CONVENTIONAL AND NUCLEAR FORCE RESTRUCTURING IN EUROPE * 6. CONFLICT AVOIDANCE MODEL * 7. GENERAL DISCUSSION OF THE WORLD LAB PROJECTS * "East-West-North-South Collaboration in Subnuclear Physics" * "Status of the World Lab in the USSR" * CLOSING SESSION

A memorial colloquium honoring Herbert L. Anderson

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

Nagle, D.E.

This paper is the result of a colloquium honoring Herbert Anderson. The paper contains memorial statements to the late Mr. Anderson and reports on; chemical analysis of the 1988 Soviet mission to Mars, Mammalian cell genetic regulation and the nature of cancer, and clean atmosphere and nuclear reactors. (JEF)

75 FR 54961 - Final Supplemental Environmental Impact Statement, Single Nuclear Unit at the Bellefonte Plant...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-09

... facility, a wind farm, a methane- gas cofiring facility, and several small solar photovoltaic facilities... maintenance of select plant systems and other regulatory compliance activities. Major buildings and plant... the plant cooling towers and the reactor, auxiliary, control, turbine, office, and service buildings...

78 FR 45984 - Yankee Atomic Electric Company, Yankee Nuclear Power Station

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-30

... Electric Company, Yankee Nuclear Power Station AGENCY: Nuclear Regulatory Commission. ACTION: Environmental... (YAEC) is the holder of Possession-Only License DPR-3 for the Yankee Nuclear Power Station (YNPS... on the site of any nuclear power reactor. In its Statement of Considerations (SOC) for the Final Rule...

78 FR 46255 - Revisions to Environmental Review for Renewal of Nuclear Power Plant Operating Licenses; Correction

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-31

... environmental effect of renewing the operating license of a nuclear power plant. This document is necessary to..., Environmental impact statement, Nuclear materials, Nuclear power plants and reactors, Reporting and... Environmental Review for Renewal of Nuclear Power Plant Operating Licenses; Correction AGENCY: Nuclear...

Final environmental impact statement. Management of commercially generated radioactive waste. Volume 1 of 3

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

Not Available

1980-10-01

This EIS analyzes the significant environmental impacts that could occur if various technologies for management and disposal of high-level and transuranic wastes from commercial nuclear power reactors were to be developed and implemented. This EIS will serve as the environmental input for the decision on which technology, or technologies, will be emphasized in further research and development activities in the commercial waste management program. The action proposed in this EIS is to (1) adopt a national strategy to develop mined geologic repositories for disposal of commercially generated high-level and transuranic radioactive waste (while continuing to examine subseabed and very deepmore » hole disposal as potential backup technologies) and (2) conduct a R and D program to develop such facilities and the necessary technology to ensure the safe long-term containment and isolation of these wastes. The Department has considered in this statement: development of conventionally mined deep geologic repositories for disposal of spent fuel from nuclear power reactors and/or radioactive fuel reprocessing wastes; balanced development of several alternative disposal methods; and no waste disposal action. This EIS reflects the public review of and comments offered on the draft statement. Included are descriptions of the characteristics of nuclear waste, the alternative disposal methods under consideration, and potential environmental impacts and costs of implementing these methods. Because of the programmatic nature of this document and the preliminary nature of certain design elements assumed in assessing the environmental consequences of the various alternatives, this study has been based on generic, rather than specific, systems. At such time as specific facilities are identified for particular sites, statements addressing site-specific aspects will be prepared for public review and comment.« less

Status of the US RERTR Program

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

Travelli, A.

1995-02-01

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

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

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

None

1958-01-15

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

The optimization of nuclear power plants operation modes in emergency situations

NASA Astrophysics Data System (ADS)

Zagrebayev, A. M.; Trifonenkov, A. V.; Ramazanov, R. N.

2018-01-01

An emergency situations resulting in the necessity for temporary reactor trip may occur at the nuclear power plant while normal operating mode. The paper deals with some of the operation c aspects of nuclear power plant operation in emergency situations and during threatened period. The xenon poisoning causes limitations on the variety of statements of the problem of calculating characteristics of a set of optimal reactor power off controls. The article show a possibility and feasibility of new sets of optimization tasks for the operation of nuclear power plants under conditions of xenon poisoning in emergency circumstances.

Neutron fluxes in test reactors

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

Youinou, Gilles Jean-Michel

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

78 FR 57538 - Proposed Waste Confidence Rule and Draft Generic Environmental Impact Statement

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-19

....nrc.gov/public-involve/public-meetings/index.cfm no later than 10 days prior to the meetings. Dated at... storage of spent nuclear fuel beyond a reactor's licensed life for operation and prior to ultimate... the proposed Waste Confidence rule and DGEIS in order to maximize public participation at these...

31 CFR 500.584 - Energy sector projects in North Korea.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Energy sector projects in North Korea... Licenses, Authorizations and Statements of Licensing Policy § 500.584 Energy sector projects in North Korea... energy sector projects in North Korea in connection with that country's transition to light-water reactor...

77 FR 47123 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-07

...-Water Reactors] Steam Generator Tubes'' [Reference 7] and NEI [Nuclear Energy Institute] 97-06, ``Steam... CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these... the required structural margins of the SG tubes for both normal and accident conditions. Nuclear...

76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-16

... Environmental Impact: Availability IX. Paperwork Reduction Act Statement X. Regulatory Analysis XI. Regulatory.... ABWR; one commenter, GE Hitachi Nuclear Energy (GEH), was against the proposed amendment to the U.S... information from both the DCD developed by GE Nuclear Energy (GE) and the DCD developed by the STPNOC. The...

78 FR 54789 - Proposed Waste Confidence Rule and Draft Generic Environmental Impact Statement

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-06

.../index.cfm no later than 10 days prior to the meetings. Dated at Rockville, Maryland, this 30th day of... storage of spent nuclear fuel beyond a reactor's licensed life for operation and prior to ultimate... the proposed Waste Confidence rule and DGEIS in order to maximize public participation at these...

78 FR 65903 - Proposed Waste Confidence Rule and Draft Generic Environmental Impact Statement

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-04

.../public-involve/public-meetings/index.cfm no later than 10 days prior to each meeting. Dated at Rockville... reactor's licensed life for operation and prior to ultimate disposal (the proposed Waste Confidence rule... documents online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html . To begin the search...

78 FR 71435 - Policy Statement on the Scenario Design Framework for Stress Testing

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-29

... Statement III. Summary of Comments A. Design of Stress Test Scenarios B. Additional Variables C. Severely... policy statement and its overall organization. A. Design of Stress Test Scenarios Commenters suggested a variety of ways for the Board to alter or improve the design of stress test scenarios, including by making...

77 FR 70124 - Policy Statement on the Scenario Design Framework for Stress Testing

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-23

... Statement on the Scenario Design Framework for Stress Testing AGENCY: Board of Governors of the Federal... Board is requesting public comment on a policy statement on the approach to scenario design for stress testing that would be used in connection with the supervisory and company-run stress tests conducted under...

78 FR 9633 - Policy Statement on the Scenario Design Framework for Stress Testing

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-11

... Statement on the Scenario Design Framework for Stress Testing AGENCY: Board of Governors of the Federal... design for stress testing that would be used in connection with the supervisory and company-run stress...) requesting public comment on a policy statement on the approach to scenario design for stress testing that...

75 FR 73134 - Florida Power and Light Company, St. Lucie Plant, Units 1 and 2; Environmental Assessment and...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-29

... proposed action does not involve any physical changes to the reactor, fuel, plant, structures, support structures, water, or land at the St. Lucie Plant, Units 1 and 2, site. The proposed action is in accordance... Impact Statement for License Renewal of Nuclear Power Plants'' (GEIS). Supplement 11 of the GEIS, issued...

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

White, J.E.; Roussin, R.W.; Gilpin, H.

A version of the CRAC2 computer code applicable for use in analyses of consequences and risks of reactor accidents in case work for environmental statements has been implemented for use on the Nuclear Regulatory Commission Data General MV/8000 computer system. Input preparation is facilitated through the use of an interactive computer program which operates on an IBM personal computer. The resulting CRAC2 input deck is transmitted to the MV/8000 by using an error-free file transfer mechanism. To facilitate the use of CRAC2 at NRC, relevant background material on input requirements and model descriptions has been extracted from four reports -more » ''Calculations of Reactor Accident Consequences,'' Version 2, NUREG/CR-2326 (SAND81-1994) and ''CRAC2 Model Descriptions,'' NUREG/CR-2552 (SAND82-0342), ''CRAC Calculations for Accident Sections of Environmental Statements, '' NUREG/CR-2901 (SAND82-1693), and ''Sensitivity and Uncertainty Studies of the CRAC2 Computer Code,'' NUREG/CR-4038 (ORNL-6114). When this background information is combined with instructions on the input processor, this report provides a self-contained guide for preparing CRAC2 input data with a specific orientation toward applications on the MV/8000. 8 refs., 11 figs., 10 tabs.« less

Putting congeniality effects into context: Investigating the role of context in attitude memory using multiple paradigms.

PubMed

Waldum, Emily R; Sahakyan, Lili

2012-05-01

In three experiments, we evaluated remembering and intentional forgetting of attitude statements that were either congruent or incongruent with participants' own political attitudes. In Experiment 1, significant directed forgetting was obtained for incongruent statements, but not for congruent statements. In addition, in the remember group, recall was better for incongruent statements than congruent statements. To explain these findings, we propose a contextual competition at retrieval hypothesis, according to which incongruent statements become more strongly associated with their episodic context during encoding than do congruent statements. At the time of retrieval, incongruent statements compete with congruent statements due to the greater amount of contextual information stored in their memory trace. We tested this hypothesis in Experiment 2 by studying free recall of congruent and incongruent statements in a mixed-pure list design. In Experiment 3, memory for incongruent and congruent statements was tested under recognition test conditions that varied in terms of how much direct retrieval of contextual details they required. Overall, the results supported the contextual competition hypothesis, and they indicate the importance of context strength in both the remembering and intentional forgetting of attitude information.

Putting congeniality effects into context: Investigating the role of context in attitude memory using multiple paradigms

PubMed Central

Waldum, Emily R; Sahakyan, Lili

2013-01-01

In three experiments, we evaluated remembering and intentional forgetting of attitude statements that were either congruent or incongruent with participants’ own political attitudes. In Experiment 1, significant directed forgetting was obtained for incongruent statements, but not for congruent statements. In addition, in the remember group, recall was better for incongruent statements than congruent statements. To explain these findings, we propose a contextual competition at retrieval hypothesis, according to which incongruent statements become more strongly associated with their episodic context during encoding than do congruent statements. At the time of retrieval, incongruent statements compete with congruent statements due to the greater amount of contextual information stored in their memory trace. We tested this hypothesis in Experiment 2 by studying free recall of congruent and incongruent statements in a mixed-pure list design. In Experiment 3, memory for incongruent and congruent statements was tested under recognition test conditions that varied in terms of how much direct retrieval of contextual details they required. Overall, the results supported the contextual competition hypothesis, and they indicate the importance of context strength in both the remembering and intentional forgetting of attitude information. PMID:23440945

KINETICS OF TREAT USED AS A TEST REACTOR

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

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

1962-05-01

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

Effect of rational and irrational statements on intensity and 'inappropriateness' of emotional distress and irrational beliefs in psychotherapy patients.

PubMed

Cramer, D; Kupshik, G

1993-09-01

Ellis's rational-emotive theory postulates that since irrational statements augment emotional distress, replacing irrational with rational statements should lessen distress. This hypothesis was tested in the initial stages of psychotherapy by having 13 and 14 clinical out-patients respectively repeat for one minute either rational or irrational statements about their major presenting psychological problem. The distinction by Ellis & Harper (1975) that 'inappropriate' emotions differ qualitatively from 'appropriate' emotions was also examined. Although the experimental intervention had no effect on a post-test measure of irrational beliefs, patients repeating rational statements had significantly lower appropriate and inappropriate negative emotions at post-test, suggesting that inappropriate emotions do not differ qualitatively from appropriate emotions and that making rational statements may lower emotional distress in patients. Patients reiterating irrational statements showed no change in emotions, implying that these kinds of irrational cognitions may have already been present.

Advanced Test Reactor Tour

ScienceCinema

Miley, Don

2017-12-21

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

TRANSURANIC STUDIES STATUS AND PROBLEM STATEMENT

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

Leuze, R E

1959-04-29

The purpose of the Transuranics Program is to develop separation processes for the transuranic elements, primarily those produced by long-term neutron irradiation of Pu/sup 239/. The program includes laboratory process development, pilot-plant process testing, processing of 10 kg of Pu/sup 239/ irradiated to greater than 99% burn-up for plutonium and americium-curium recovery, and processing the reirradiated plutonium and americium-curium fractions. The proposed method for processing highly irradiated plutonium is: (1) plutonium-aluminum alloy dissolution in HNO/sub 3/; (2) plutonium recovery by TBP extraction; (3) americium, curium, and rare-earth extraction by TBP from neutral nitrate solution; (4) partial rare-earth removal (primarily lanthanum)more » by americium-curium extraction into 100% TBP from 15M HNO/sub 3/; (5) additional rare-earth removal by extraction in 0.48M mono-2-ethylhexylphosphoric acid from 12M HCl; and (6) americium-curium purification by chloride anion exchange. Processing through the 100% TBP, 15M HNO/sub 3/ cycle can be carried out in the Power Reactor Fuel Reprocessing Pilot Plant. New facilities are proposed 15M HNO/ sub 3/ cycle can be carried out in the Power Reactor Fuel Reprocessing Pilot Plant. New facilities are proposed for laboratory process development studies and the final processing of the transplutonic elements. (auth)« less

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

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

2009-07-01

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

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

NASA Astrophysics Data System (ADS)

Carlson, William F.; Bitten, Ernest J.

1993-01-01

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

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

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

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

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

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

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

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

Summary of NR Program Prometheus Efforts

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

J Ashcroft; C Eshelman

2006-02-08

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

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

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

Sienicki, James J.; Grandy, Christopher

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

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

NASA Technical Reports Server (NTRS)

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

2010-01-01

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

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

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

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

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

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

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

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

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

NASA Astrophysics Data System (ADS)

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

1981-09-01

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

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

NASA Technical Reports Server (NTRS)

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

1981-01-01

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

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

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

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

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

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

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

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

Lessons Learned about Liquid Metal Reactors from FFTF Experience

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

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

2016-09-20

The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens.more » In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports« less

Site Environmental Report for Calendar Year 2007. DOE Operations at The Boeing Company, Santa Susana Field Laboratory, Area IV

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

Liu, Ning; Rutherford, Phil; Lenox, Art

2008-09-30

This Annual Site Environmental Report (ASER) for 2007 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder components. All nuclear work was terminated in 1988; all subsequentmore » radiological work has been directed toward decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. In May 2007, the D&D operations in Area IV were suspended until DOE completes the SSFL Area IV Environmental Impact Statement (EIS). The environmental monitoring programs were continued throughout the year. Results of the radiological monitoring program for the calendar year 2007 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling. All radioactive wastes are processed for disposal at DOE disposal sites and/or other licensed sites approved by DOE for radioactive waste disposal. No liquid radioactive wastes were released into the environment in 2007.« less

A Comparison of Measured and Calculated Air-Transported Radiation from a Fast. Unshielded Nuclear Reactor.

DTIC Science & Technology

1980-12-01

UNSHIELDED NUCLEAR REACTOR by H. A. Robitaille and B. E. Hoffarth D T ICS ELECTE SEP 1 5 i9813 C- LA . DISTRIBUTION STATEMENT A O PROJECT NO. 1 DECEMBER 1980...et de neutrons b diff6rentes distances jusqu’ 1100 m~tres du r6acteur ncldaire & neutrons rapides de la Pulse Radiation Division de la U.S. Army...u)UOU 𔃽N~nl4 N0in0 -CuD z f- 000 LUU ’4 :1O 00 w4 w4 w (u3)uU3 𔃽N~n-A NO.Ln3 rc ino OQQ z ca co, 0 0- C ) LU c C. M CU m CD 0 LA (u3uOU 𔃽N~n1.4No

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

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

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

Eddy Current Flow Measurements in the FFTF

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

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

2017-02-02

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

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

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

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

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

77 FR 38751 - Codification of Animal Testing Policy

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-29

... Animal Testing Policy AGENCY: Consumer Product Safety Commission. ACTION: Proposed Statement of Policy on Animal Testing SUMMARY: The Consumer Product Safety Commission (CPSC or Commission) proposes to codify its statement of policy on animal testing, as amended, which was previously published in the Federal...

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

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

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

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

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

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

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

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

1993-01-01

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

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

ScienceCinema

None

2018-01-16

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

12 CFR Appendix A to Part 252 - Policy Statement on the Scenario Design Framework for Stress Testing

Code of Federal Regulations, 2014 CFR

2014-01-01

... 12 Banks and Banking 4 2014-01-01 2014-01-01 false Policy Statement on the Scenario Design... YY) Pt. 252, App. A Appendix A to Part 252—Policy Statement on the Scenario Design Framework for... (stress test rules) implementing section 165(i) of the Dodd-Frank Wall Street Reform and Consumer...

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

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

1974-01-01

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

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

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

1974-01-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-17

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

A document review to characterize Atomic International SNAP fuels shipped to INEL 1966--1973

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

Wahnschaffe, S.D.; Lords, R.E.; Kneff, D.W.

1995-09-01

This report provides the results of a document search and review study to obtain information on the spent fuels for the following six Nuclear Auxiliary Power (SNAP) reactor cores now stored at the Idaho National Engineering Laboratory (INEL): SNAP-2 Experimental Reactor, SNAP-2 Development Reactor, SNAP-10A Ground Test Reactor, SNAP-8 Experimental Reactor, SNAP-8 Development Reactor, and Shield Test Reactor. The report also covers documentation on SNAP fuel materials from four in-pile materials tests: NAA-82-1, NAA-115-2, NAA-117-1, and NAA-121. Pieces of these fuel materials are also stored at INEL as part of the SNAP fuel shipments.

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

NASA Technical Reports Server (NTRS)

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

1972-01-01

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

40 CFR 790.99 - Statement of financial responsibility.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 31 2010-07-01 2010-07-01 true Statement of financial responsibility... From Test Rules § 790.99 Statement of financial responsibility. Each applicant for an exemption shall submit the following sworn statement with his or her application: I understand that if this application...

Operators in the Plum Brook Reactor Facility Control Room

NASA Image and Video Library

1970-03-21

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

Space reactor fuel element testing in upgraded TREAT

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

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

1993-01-14

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

Space reactor fuel element testing in upgraded TREAT

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

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

1993-05-01

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

Oxidation of aluminum alloy cladding for research and test reactor fuel

NASA Astrophysics Data System (ADS)

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

2008-08-01

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

SNAP 10A FS-3 reactor performance

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

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

1966-08-15

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

78 FR 64153 - Policy Statement on the Principles for Development and Distribution of Annual Stress Test Scenarios

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-28

.... OCC-2012-0016] Policy Statement on the Principles for Development and Distribution of Annual Stress... in developing and distributing the stress test scenarios for the annual stress test required by the... by the Annual Stress Test final rule (Stress Test Rule) published on October 9, 2012. Under the...

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

NASA Technical Reports Server (NTRS)

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

1972-01-01

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

Exploratory evaluation of ceramics for automobile thermal reactors

NASA Technical Reports Server (NTRS)

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

1972-01-01

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

Lyin’ Eyes: Ocular-motor Measures of Reading Reveal Deception

PubMed Central

Cook, Anne E.; Hacker, Douglas J.; Webb, Andrea K.; Osher, Dahvyn; Kristjansson, Sean; Woltz, Dan J.; Kircher, John C.

2013-01-01

Our goal was to evaluate an alternative to current methods for detecting deception in security screening contexts. We evaluated a new cognitive-based test of deception that measured participants’ ocular-motor responses (pupil responses and reading behaviors) while they read and responded to statements on a computerized questionnaire. In Experiment 1, participants from a university community were randomly assigned to either a “guilty” group that committed one of two mock crimes or an “innocent” group that only learned about the crime. Participants then reported for testing, where they completed the computer-administered questionnaire that addressed their possible involvement in the crimes. Experiment 2 also manipulated participants’ incentive to pass the test and difficulty of statements on the test. In both experiments, guilty participants had increased pupil responses to statements answered deceptively; however, they spent less time fixating on, reading, and re-reading those statements than statements answered truthfully. These ocular-motor measures were optimally weighted in a discrimination function that correctly classified 85% of participants as either guilty or innocent. Findings from Experiment 2 indicated that group discrimination was improved with greater incentives to pass the test and the use of statements with simple syntax. The present findings suggest that two cognitive processes are involved in deception -- vigilance and strategy -- and that these processes are reflected in different ocular-motor measures. The ocular-motor test reported here represents a new approach to detecting deception that may fill an important need in security screening contexts. PMID:22545928

Valx: A system for extracting and structuring numeric lab test comparison statements from text

PubMed Central

Hao, Tianyong; Liu, Hongfang; Weng, Chunhua

2017-01-01

Objectives To develop an automated method for extracting and structuring numeric lab test comparison statements from text and evaluate the method using clinical trial eligibility criteria text. Methods Leveraging semantic knowledge from the Unified Medical Language System (UMLS) and domain knowledge acquired from the Internet, Valx takes 7 steps to extract and normalize numeric lab test expressions: 1) text preprocessing, 2) numeric, unit, and comparison operator extraction, 3) variable identification using hybrid knowledge, 4) variable - numeric association, 5) context-based association filtering, 6) measurement unit normalization, and 7) heuristic rule-based comparison statements verification. Our reference standard was the consensus-based annotation among three raters for all comparison statements for two variables, i.e., HbA1c and glucose, identified from all of Type 1 and Type 2 diabetes trials in ClinicalTrials.gov. Results The precision, recall, and F-measure for structuring HbA1c comparison statements were 99.6%, 98.1%, 98.8% for Type 1 diabetes trials, and 98.8%, 96.9%, 97.8% for Type 2 Diabetes trials, respectively. The precision, recall, and F-measure for structuring glucose comparison statements were 97.3%, 94.8%, 96.1% for Type 1 diabetes trials, and 92.3%, 92.3%, 92.3% for Type 2 diabetes trials, respectively. Conclusions Valx is effective at extracting and structuring free-text lab test comparison statements in clinical trial summaries. Future studies are warranted to test its generalizability beyond eligibility criteria text. The open-source Valx enables its further evaluation and continued improvement among the collaborative scientific community. PMID:26940748

Valx: A System for Extracting and Structuring Numeric Lab Test Comparison Statements from Text.

PubMed

Hao, Tianyong; Liu, Hongfang; Weng, Chunhua

2016-05-17

To develop an automated method for extracting and structuring numeric lab test comparison statements from text and evaluate the method using clinical trial eligibility criteria text. Leveraging semantic knowledge from the Unified Medical Language System (UMLS) and domain knowledge acquired from the Internet, Valx takes seven steps to extract and normalize numeric lab test expressions: 1) text preprocessing, 2) numeric, unit, and comparison operator extraction, 3) variable identification using hybrid knowledge, 4) variable - numeric association, 5) context-based association filtering, 6) measurement unit normalization, and 7) heuristic rule-based comparison statements verification. Our reference standard was the consensus-based annotation among three raters for all comparison statements for two variables, i.e., HbA1c and glucose, identified from all of Type 1 and Type 2 diabetes trials in ClinicalTrials.gov. The precision, recall, and F-measure for structuring HbA1c comparison statements were 99.6%, 98.1%, 98.8% for Type 1 diabetes trials, and 98.8%, 96.9%, 97.8% for Type 2 diabetes trials, respectively. The precision, recall, and F-measure for structuring glucose comparison statements were 97.3%, 94.8%, 96.1% for Type 1 diabetes trials, and 92.3%, 92.3%, 92.3% for Type 2 diabetes trials, respectively. Valx is effective at extracting and structuring free-text lab test comparison statements in clinical trial summaries. Future studies are warranted to test its generalizability beyond eligibility criteria text. The open-source Valx enables its further evaluation and continued improvement among the collaborative scientific community.

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

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

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

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

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

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-22

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

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

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

2006-01-01

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

Preliminary plan for testing a thermionic reactor in the Plum Brook Space Power Facility

NASA Technical Reports Server (NTRS)

Haley, F. A.

1972-01-01

A preliminary plan is presented for testing a thermionic reactor in the Plum Brook Space Power Facility (SPF). A technical approach, cost estimate, manpower estimate, and schedule are presented to cover a 2 year full power reactor test.

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

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

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

2006-10-01

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

Preliminary Options Assessment of Versatile Irradiation Test Reactor

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

Sen, Ramazan Sonat

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

Redefining Success: Applying Lessons in Nuclear Diplomacy from North Korea to Iran (INSS Strategic Perspectives, Number 1, September 2010)

DTIC Science & Technology

2010-09-01

DISTRIBUTION/AVAILABILITY STATEMENT Approved for public release; distribution unlimited 13 . SUPPLEMENTARY NOTES 14. ABSTRACT 15. SUBJECT TERMS...consideration of a U.S. military strike), peacefully return North Korea to the NPT and freeze its nuclear program 19 months later. 13 In the process...lifted the nuclear freeze and in January 2003 withdrew from 13 Redefining Success the NPT, citing two main reasons: U.S. failure to honor the reactor

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

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

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

2016-01-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

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

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

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

2016-03-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

Space station prototype Sabatier reactor design verification testing

NASA Technical Reports Server (NTRS)

Cusick, R. J.

1974-01-01

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

Points to Consider: Ethical, Legal, and Psychosocial Implications of Genetic Testing in Children and Adolescents

PubMed Central

Botkin, Jeffrey R.; Belmont, John W.; Berg, Jonathan S.; Berkman, Benjamin E.; Bombard, Yvonne; Holm, Ingrid A.; Levy, Howard P.; Ormond, Kelly E.; Saal, Howard M.; Spinner, Nancy B.; Wilfond, Benjamin S.; McInerney, Joseph D.

2015-01-01

In 1995, the American Society of Human Genetics (ASHG) and American College of Medical Genetics and Genomics (ACMG) jointly published a statement on genetic testing in children and adolescents. In the past 20 years, much has changed in the field of genetics, including the development of powerful new technologies, new data from genetic research on children and adolescents, and substantial clinical experience. This statement represents current opinion by the ASHG on the ethical, legal, and social issues concerning genetic testing in children. These recommendations are relevant to families, clinicians, and investigators. After a brief review of the 1995 statement and major changes in genetic technologies in recent years, this statement offers points to consider on a broad range of test technologies and their applications in clinical medicine and research. Recommendations are also made for record and communication issues in this domain and for professional education. PMID:26140447

CRITICAL EXPERIMENT TANK (CET) REACTOR HAZARDS SUMMARY

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

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

1961-03-31

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

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

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

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

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

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

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

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

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

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

JPRS Report, Science & Technology, China: Energy.

DTIC Science & Technology

1992-03-30

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

HEDL FACILITIES CATALOG 400 AREA

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

MAYANCSIK BA

1987-03-01

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

Tory II-A: a nuclear ramjet test reactor

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

Hadley, J.W.

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

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

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

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

77 FR 51528 - Notice of Public Hearings for the Draft Environmental Impact Statement for Outdoor Research...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-24

... Impact Statement for Outdoor Research, Development, Test and Evaluation Activities, Naval Surface Warfare..., development, test and evaluation (RDT&E) activities within the Potomac River Test Range (PRTR) complex... current, baseline RDT&E mission requirements for activities that have the potential to affect human health...

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

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

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

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

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

PubMed Central

Fox, Peter; Suidan, Makram T.

1990-01-01

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

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

PubMed

Fox, P; Suidan, M T

1990-04-01

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

Nuclear Reactors. Revised.

ERIC Educational Resources Information Center

Hogerton, John F.

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

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

NASA Astrophysics Data System (ADS)

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

2004-02-01

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

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

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

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

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

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

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-10

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

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

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

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

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

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

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

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

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

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

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

ERIC Educational Resources Information Center

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

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

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

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

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

2009-09-15

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

Dismantling of Loop-Type Channel Equipment of MR Reactor in NRC 'Kurchatov Institute' - 13040

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

Volkov, Victor; Danilovich, Alexey; Zverkov, Yuri

2013-07-01

In 2009 the project of decommissioning of MR and RTF reactors was developed and approved by the Expert Authority of the Russian Federation (Gosexpertiza). The main objective of the decommissioning works identified in this project: - complete dismantling of reactor equipment and systems; - decontamination of reactor premises and site in accordance with the established sanitary and hygienic standards. At the preparatory stage (2008-2010) of the project the following works were executed: loop-type channels' dismantling in the storage pool; experimental fuel assemblies' removal from spent fuel repositories in the central hall; spent fuel assembly removal from the liquid-metal-cooled loop-type channelmore » of the reactor core and its placement into the SNF repository; and reconstruction of engineering support systems to the extent necessary for reactor decommissioning. The project assumes three main phases of dismantling and decontamination: - dismantling of equipment/pipelines of cooling circuits and loop-type channels, and auxiliary reactor equipment (2011-2012); - dismantling of equipment in underground reactor premises and of both MR and RTF in-vessel devices (2013-2014); - decontamination of reactor premises; rehabilitation of the reactor site; final radiation survey of reactor premises, loop-type channels and site; and issuance of the regulatory authorities' de-registration statement (2015). In 2011 the decommissioning license for the two reactors was received and direct MR decommissioning activities started. MR primary pipelines and loop-type facilities situated in the underground reactor hall were dismantled. Works were also launched to dismantle the loop-type channels' equipment in underground reactor premises; reactor buildings were reconstructed to allow removal of dismantled equipment; and the MR/RTF decommissioning sequence was identified. In autumn 2011 - spring 2012 results of dismantling activities performed are: - equipment from underground rooms (No. 66, 66A, 66B, 72, 64, 63) - as well as from water and gas loop corridors - was dismantled, with the total radwaste weight of 53 tons and the total removed activity of 5,0 x 10{sup 10} Bq; - loop-type channel equipment from underground reactor hall premises was dismantled; - 93 loop-type channels were characterized, chopped and removed, with radwaste of 2.6 x 10{sup 13} Bq ({sup 60}Co) and 1.5 x 10{sup 13} Bq ({sup 137}Cs) total activity removed from the reactor pool, fragmented and packaged. Some of this waste was placed into the high-level waste (HLW) repository of the Center. Dismantling works were executed with application of remotely operated mechanisms, which promoted decrease of radiation impact on the personnel. The average individual dose for the personnel was 1.9 mSv/year in 2011, and the collective dose is estimated as 0.0605 man x Sv/year. (authors)« less

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

PubMed Central

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

1978-01-01

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

10 CFR 52.167 - Issuance of manufacturing license.

Code of Federal Regulations, 2010 CFR

2010-01-01

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

DEVICE FOR TREATING MATERIALS

DOEpatents

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

1959-02-17

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

In-reactor performance of LWR-type tritium target rods

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

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

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

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Tell Me Everything You Discussed: Children's Memory for Dyadic Conversations after a 1-Week or a 3-Week Delay.

PubMed

Lawson, Monica; London, Kamala

2015-08-01

In child abuse investigations, children are often asked to recount previous conversations related to the allegations (i.e., "conversational testimony"). To explore children's ability to provide conversational testimony, we staged a semi-structured novel dyadic conversation between an adult researcher and 8-year-old children (n = 90). Children's gist recall and recognition memory for their own statements, their conversational partner's statements, and question-answer pairs were tested after either a 1-week or a 3-week delay. The results revealed that children recounted a minority of the conversation, although children recalled more after a short delay (7%) than after a long delay (4%). A majority of children's free recall statements were accurate (68%); however, approximately one-third of their free recall statements were incorrect. Children almost exclusively recounted their own statements, and rarely recalled any of the adult's statements or the question-answer pairs during free recall. Reports of the adult's statements and question-answer pairs increased with cued recall questioning, but remained minimal. During recognition testing, children were able to distinguish between true and false recognition items for their own statements and the adult's statements, but performed at chance level on recognition items concerning question-answer pairs. Forensic implications of the results are discussed. Copyright © 2015 John Wiley & Sons, Ltd.

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

NASA Technical Reports Server (NTRS)

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

2007-01-01

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

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2012 CFR

2012-01-01

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

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2013 CFR

2013-01-01

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

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

PubMed

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

2008-02-01

Dyed microspheres have been developed as a new method for validation of ultraviolet (UV) reactor systems. When properly applied, dyed microspheres allow measurement of the UV dose distribution delivered by a photochemical reactor for a given operating condition. Prior to this research, dyed microspheres had only been applied to a bench-scale UV reactor. The goal of this research was to extend the application of dyed microspheres to large-scale reactors. Dyed microsphere tests were conducted on two prototype large-scale UV reactors at the UV Validation and Research Center of New York (UV Center) in Johnstown, NY. All microsphere tests were conducted under conditions that had been used previously in biodosimetry experiments involving two challenge bacteriophage: MS2 and Qbeta. Numerical simulations based on computational fluid dynamics and irradiance field modeling were also performed for the same set of operating conditions used in the microspheres assays. Microsphere tests on the first reactor illustrated difficulties in sample collection and discrimination of microspheres against ambient particles. Changes in sample collection and work-up were implemented in tests conducted on the second reactor that allowed for improvements in microsphere capture and discrimination against the background. Under these conditions, estimates of the UV dose distribution from the microspheres assay were consistent with numerical simulations and the results of biodosimetry, using both challenge organisms. The combined application of dyed microspheres, biodosimetry, and numerical simulation offers the potential to provide a more in-depth description of reactor performance than any of these methods individually, or in combination. This approach also has the potential to substantially reduce uncertainties in reactor validation, thereby leading to better understanding of reactor performance, improvements in reactor design, and decreases in reactor capital and operating costs.

Improving perceptions of empathy in patients undergoing low-yield computerized tomographic imaging in the emergency department.

PubMed

Lin, Michelle P; Probst, Marc A; Puskarich, Michael A; Dehon, Erin; Kuehl, Damon R; Wang, Ralph C; Hess, Erik P; Butler, Katie; Runyon, Michael S; Wang, Hao; Courtney, D Mark; Muckley, Brandon; Hobgood, Cherri D; Hall, Cassandra L; Kline, Jeffrey A

2018-04-01

We assessed emergency department (ED) patient perceptions of how physicians can improve their language to determine patient preferences for 11 phrases to enhance physician empathy toward the goal of reducing low-value advanced imaging. Multi-center survey study of low-risk ED patients undergoing computerized tomography (CT) scanning. We enroled 305 participants across nine sites. The statement "I have carefully considered what you told me about what brought you here today" was most frequently rated as important (88%). The statement "I have thought about the cost of your medical care to you today" was least frequently rated as important (59%). Participants preferred statements indicating physicians had considered their "vital signs and physical examination" (86%), "past medical history" (84%), and "what prior research tells me about your condition" (79%). Participants also valued statements conveying risks of testing, including potential kidney injury (78%) and radiation (77%). The majority of phrases were identified as important. Participants preferred statements conveying cognitive reassurance, medical knowledge and risks of testing. Our findings suggest specific phrases have the potential to enhance ED patient perceptions of physician empathy. Further research is needed to determine whether statements to convey empathy affect diagnostic testing rates. Copyright © 2017 Elsevier B.V. All rights reserved.

Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

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

Tsai, H.; Gomes, I.C.; Smith, D.L.

1998-09-01

The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2011 CFR

2011-01-01

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

TREAT Reactor Control and Protection System

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

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS).more » The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.« less

American Society of Clinical Oncology Policy Statement Update: Genetic and Genomic Testing for Cancer Susceptibility.

PubMed

Robson, Mark E; Bradbury, Angela R; Arun, Banu; Domchek, Susan M; Ford, James M; Hampel, Heather L; Lipkin, Stephen M; Syngal, Sapna; Wollins, Dana S; Lindor, Noralane M

2015-11-01

The American Society of Clinical Oncology (ASCO) has long affirmed that the recognition and management of individuals with an inherited susceptibility to cancer are core elements of oncology care. ASCO released its first statement on genetic testing in 1996 and updated that statement in 2003 and 2010 in response to developments in the field. In 2014, the Cancer Prevention and Ethics Committees of ASCO commissioned another update to reflect the impact of advances in this area on oncology practice. In particular, there was an interest in addressing the opportunities and challenges arising from the application of massively parallel sequencing-also known as next-generation sequencing-to cancer susceptibility testing. This technology introduces a new level of complexity into the practice of cancer risk assessment and management, requiring renewed effort on the part of ASCO to ensure that those providing care to patients with cancer receive the necessary education to use this new technology in the most effective, beneficial manner. The purpose of this statement is to explore the challenges of new and emerging technologies in cancer genetics and provide recommendations to ensure their optimal deployment in oncology practice. Specifically, the statement makes recommendations in the following areas: germline implications of somatic mutation profiling, multigene panel testing for cancer susceptibility, quality assurance in genetic testing, education of oncology professionals, and access to cancer genetic services. © 2015 by American Society of Clinical Oncology.

77 FR 68047 - Policy Statement on the Principles for Development and Distribution of Annual Stress Test Scenarios

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-15

.... OCC-2012-0016] Policy Statement on the Principles for Development and Distribution of Annual Stress... and factors to be used by the OCC in development and distributing the stress test scenarios for the annual stress test required by the Dodd- Frank Wall Street Reform and Consumer Protection Act of 2010 as...

Laboratory testing of extravascular body fluids in Croatia: a survey of the Working group for extravascular body fluids of the Croatian Society of Medical Biochemistry and Laboratory Medicine.

PubMed

Kopcinovic, Lara Milevoj; Vogrinc, Zeljka; Kocijan, Irena; Culej, Jelena; Aralica, Merica; Jokic, Anja; Antoncic, Dragana; Bozovic, Marija

2016-10-15

We hypothesized that extravascular body fluid (EBF) analysis in Croatia is not harmonized and aimed to investigate preanalytical, analytical and postanalytical procedures used in EBF analysis in order to identify key aspects that should be addressed in future harmonization attempts. An anonymous online survey created to explore laboratory testing of EBF was sent to secondary, tertiary and private health care Medical Biochemistry Laboratories (MBLs) in Croatia. Statements were designed to address preanalytical, analytical and postanalytical procedures of cerebrospinal, pleural, peritoneal (ascites), pericardial, seminal, synovial, amniotic fluid and sweat. Participants were asked to declare the strength of agreement with proposed statements using a Likert scale. Mean scores for corresponding separate statements divided according to health care setting were calculated and compared. The survey response rate was 0.64 (58 / 90). None of the participating private MBLs declared to analyse EBF. We report a mean score of 3.45 obtained for all statements evaluated. Deviations from desirable procedures were demonstrated in all EBF testing phases. Minor differences in procedures used for EBF analysis comparing secondary and tertiary health care MBLs were found. The lowest scores were obtained for statements regarding quality control procedures in EBF analysis, participation in proficiency testing programmes and provision of interpretative comments on EBF's test reports. Although good laboratory EBF practice is present in Croatia, procedures for EBF analysis should be further harmonized to improve the quality of EBF testing and patient safety.

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

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

Not Available

1992-07-01

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

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

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

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

2015-04-01

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

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

NONE

Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of researchmore » reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)« less

OAST Space Theme Workshop. Volume 3: Working group summary. 6: Power (P-2). A. Statement. B. Technology needs (form 1). C. Priority assessment (form 2)

NASA Technical Reports Server (NTRS)

1976-01-01

Power requirements for the multipurpose space power platform, for space industrialization, SETI, the solar system exploration facility, and for global services are assessed for various launch dates. Priorities and initiatives for the development of elements of space power systems are described for systems using light power input (solar energy source) or thermal power input, (solar, chemical, nuclear, radioisotopes, reactors). Systems for power conversion, power processing, distribution and control are likewise examined.

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2013 CFR

2013-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2011 CFR

2011-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2012 CFR

2012-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2014 CFR

2014-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

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

NASA Technical Reports Server (NTRS)

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

2016-01-01

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

A liquid-metal filling system for pumped primary loop space reactors

NASA Astrophysics Data System (ADS)

Crandall, D. L.; Reed, W. C.

Some concepts for the SP-100 space nuclear power reactor use liquid metal as the primary coolant in a pumped loop. Prior to filling ground engineering test articles or reactor systems, the liquid metal must be purified and circulated through the reactor primary system to remove contaminants. If not removed, these contaminants enhance corrosion and reduce reliability. A facility was designed and built to support Department of Energy Liquid Metal Fast Breeder Reactor tests conducted at the Idaho National Engineering Laboratory. This test program used liquid sodium to cool nuclear fuel in in-pile experiments; thus, a system was needed to store and purify sodium inventories and fill the experiment assemblies. This same system, with modifications and potential changeover to lithium or sodium-potassium (NaK), can be used in the Space Nuclear Power Reactor Program. This paper addresses the requirements, description, modifications, operation, and appropriateness of using this liquid-metal system to support the SP-100 space reactor program.

10 CFR 171.3 - Scope.

Code of Federal Regulations, 2010 CFR

2010-01-01

... holding an operating license for a power reactor, test reactor or research reactor issued under part 50 of... authorizes operation of a power reactor. The regulations in this part also apply to any person holding a...

Cognitive restructuring and exposure with a compulsive washer: a test of reciprocal inhibition.

PubMed

Giles, T R

1982-09-01

Reciprocal inhibition was tested with a compulsive washer using the affect elicited by rational self-statements as the response competitor to anxiety. Three conditions, CS alone, CS plus irrational self-statements, and CS plus rational self-statements, were randomly and continuously repeated over six trials. The CS was presented both imaginally and in vivo. It was hypothesized: (1) that subjective anxiety in all conditions would substantially diminish by the experiment's end; (2) that the rational self-statement condition would reduce (inhibit) subjective anxiety to a significantly greater degree than the other two conditions; and (3) that significant anxiety reduction would occur only during the rational restructuring trials. All three hypotheses were supported by the results.

Advanced Instrumentation for Transient Reactor Testing

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

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

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

REACTOR SERVICE BUILDING, TRA635, CONTEXTUAL VIEW DURING CONSTRUCTION. CAMERA IS ...

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

REACTOR SERVICE BUILDING, TRA-635, CONTEXTUAL VIEW DURING CONSTRUCTION. CAMERA IS ATOP MTR BUILDING AND LOOKING SOUTHERLY. FOUNDATION AND DRAINS ARE UNDER CONSTRUCTION. THE BUILDING WILL BUTT AGAINST CHARGING FACE OF PLUG STORAGE BUILDING. HOT CELL BUILDING, TRA-632, IS UNDER CONSTRUCTION AT TOP CENTER OF VIEW. INL NEGATIVE NO. 8518. Unknown Photographer, 8/25/1953 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Reactor transient control in support of PFR/TREAT TUCOP experiments

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

Burrows, D.R.; Larsen, G.R.; Harrison, L.J.

1984-01-01

Unique energy deposition and experiment control requirements posed bythe PFR/TREAT series of transient undercooling/overpower (TUCOP) experiments resulted in equally unique TREAT reactor operations. New reactor control computer algorithms were written and used with the TREAT reactor control computer system to perform such functions as early power burst generation (based on test train flow conditions), burst generation produced by a step insertion of reactivity following a controlled power ramp, and shutdown (SCRAM) initiators based on both test train conditions and energy deposition. Specialized hardware was constructed to simulate test train inputs to the control computer system so that computer algorithms couldmore » be tested in real time without irradiating the experiment.« less

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

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

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

HOT CELL BUILDING, TRA632. CONTEXTUAL VIEW ALONG WALLEYE AVENUE, CAMERA ...

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

HOT CELL BUILDING, TRA-632. CONTEXTUAL VIEW ALONG WALLEYE AVENUE, CAMERA FACING EASTERLY. HOT CELL BUILDING IS AT CENTER LEFT OF VIEW; THE LOW-BAY PROJECTION WITH LADDER IS THE TEST TRAIN ASSEMBLY FACILITY, ADDED IN 1968. MTR BUILDING IS IN LEFT OF VIEW. HIGH-BAY BUILDING AT RIGHT IS THE ENGINEERING TEST REACTOR BUILDING, TRA-642. INL NEGATIVE NO. HD46-32-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins

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

Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara

2005-02-06

Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an earlymore » prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore called 'HPR-1'.« less

Environmental assessment of SP-100 ground engineering system test site: Hanford Site, Richland, Washington

NASA Astrophysics Data System (ADS)

1988-12-01

The US Department of Energy (DOE) proposes to modify an existing reactor containment building (decommissioned Plutonium Recycle Test Reactor (PRTR) 309 Building) to provide ground test capability for the prototype SP-100 reactor. The 309 Building (Figure 1.1) is located in the 300 Area on the Hanford Site in Washington State. The National Environmental Policy Act (NEPA) requires that Federal agencies assess the potential impacts that their actions may have on the environment. This Environmental Assessment describes the consideration given to environmental impacts during reactor concept and test site selection, examines the environmental effects of the DOE proposal to ground test the nuclear subsystem, describes alternatives to the proposed action, and examines radiological risks of potential SP-100 use in space.

A new self-report inventory of dyslexia for students: criterion and construct validity.

PubMed

Tamboer, Peter; Vorst, Harrie C M

2015-02-01

The validity of a Dutch self-report inventory of dyslexia was ascertained in two samples of students. Six biographical questions, 20 general language statements and 56 specific language statements were based on dyslexia as a multi-dimensional deficit. Dyslexia and non-dyslexia were assessed with two criteria: identification with test results (Sample 1) and classification using biographical information (both samples). Using discriminant analyses, these criteria were predicted with various groups of statements. All together, 11 discriminant functions were used to estimate classification accuracy of the inventory. In Sample 1, 15 statements predicted the test criterion with classification accuracy of 98%, and 18 statements predicted the biographical criterion with classification accuracy of 97%. In Sample 2, 16 statements predicted the biographical criterion with classification accuracy of 94%. Estimations of positive and negative predictive value were 89% and 99%. Items of various discriminant functions were factor analysed to find characteristic difficulties of students with dyslexia, resulting in a five-factor structure in Sample 1 and a four-factor structure in Sample 2. Answer bias was investigated with measures of internal consistency reliability. Less than 20 self-report items are sufficient to accurately classify students with and without dyslexia. This supports the usefulness of self-assessment of dyslexia as a valid alternative to diagnostic test batteries. Copyright © 2015 John Wiley & Sons, Ltd.

An evaluation of alloys and coatings for use in automobile thermal reactors

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Oldrieve, R. E.

1974-01-01

Several candidate alloys and coatings were evaluated for use in automobile thermal reactors. Full-size reactors of the candidate materials were analyzed in cyclic engine dynamometer tests with peak temperature of 1900 F (1040 C). Two developmental ferritic iron alloys GE1541 and NASA-18T - exhibited the best overall performance lasting at least 60% of the life of the test engine. Four of the alloys evaluated warrant consideration for reactor use. They include GE1541, Armco 18 SR, NASA-18T, and Inconel 601. None of the commercial coating substrate combinations evaluated warrant consideration for reactor use.-

Evaluation of alloys and coatings for use in automobile thermal reactors

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Oldrieve, R. E.

1974-01-01

Several candidate alloys and coatings were evaluated for use in automobile thermal reactors. Full-size reactors of the candidate materials were evaluated in cyclic engine dynamometer tests with a peak temperature of 1040 C (1900 F). Two developmental ferritic-iron alloys, GE-1541 and NASA-18T, exhibited the best overall performance by lasting at least 60 percent of the life of test engine. Four of the alloys evaluated warrant consideration for reactor use. They are GE-1541, Armco 18 SR, NASA-18T, and Inconel 601. None of the commercial coating substrate combinations evaluated warrant consideration for reactor use.

President Clinton's Statement on the Comprehensive Nuclear Test Ban Treaty

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

Clinton, Bill

This is video footage of President Clinton delivering a statement to the press on signing the Comprehensive Nuclear Test Ban Treaty and answering press pool questions before departing Kansas City, Missouri. This footage is official public record produced by the White House Television (WHTV) crew, provided by the Clinton Presidential Library.

Testing Some Notions of the Fact/Opinion Distinction in Libel.

ERIC Educational Resources Information Center

Cohen, Jeremy; And Others

A study of reader response to newspaper articles in a defamatory context tested: (1) the judicial assumption that the macro-environment in which statements appear is important to a reader's distinguishing between fact and opinion; (2) the possibility that a byline may influence a reader's characterization of statements; and (3) the idea that…

49 CFR 173.476 - Approval of special form Class 7 (radioactive) materials.

Code of Federal Regulations, 2014 CFR

2014-10-01

... any tests, demonstrating that the special form material meets the requirements of § 173.469. An IAEA... both physical and chemical states; (2) A detailed statement of the capsule design and dimensions..., and methods of construction; (3) A statement of the tests that have been made and their results; or...

49 CFR 173.476 - Approval of special form Class 7 (radioactive) materials.

Code of Federal Regulations, 2012 CFR

2012-10-01

... tests, demonstrating that the special form material meets the requirements of § 173.469. An IAEA... both physical and chemical states; (2) A detailed statement of the capsule design and dimensions..., and methods of construction; (3) A statement of the tests that have been made and their results; or...

49 CFR 173.476 - Approval of special form Class 7 (radioactive) materials.

Code of Federal Regulations, 2013 CFR

2013-10-01

... tests, demonstrating that the special form material meets the requirements of § 173.469. An IAEA... both physical and chemical states; (2) A detailed statement of the capsule design and dimensions..., and methods of construction; (3) A statement of the tests that have been made and their results; or...

49 CFR 173.476 - Approval of special form Class 7 (radioactive) materials.

Code of Federal Regulations, 2010 CFR

2010-10-01

... tests, demonstrating that the special form material meets the requirements of § 173.469. An IAEA... both physical and chemical states; (2) A detailed statement of the capsule design and dimensions..., and methods of construction; (3) A statement of the tests that have been made and their results; or...

49 CFR 173.476 - Approval of special form Class 7 (radioactive) materials.

Code of Federal Regulations, 2011 CFR

2011-10-01

... tests, demonstrating that the special form material meets the requirements of § 173.469. An IAEA... both physical and chemical states; (2) A detailed statement of the capsule design and dimensions..., and methods of construction; (3) A statement of the tests that have been made and their results; or...

The effect of catalyst length and downstream reactor distance on catalytic combustor performance

NASA Technical Reports Server (NTRS)

Anderson, D.

1980-01-01

A study was made to determine the effects on catalytic combustor performance which resulted from independently varying the length of a catalytic reactor and the length available for gas-phase reactions downstream of the catalyst. Monolithic combustion catalysts from three manufacturers were tested in a combustion test rig with no. 2 diesel fuel. Catalytic reactor lengths of 2.5 and 5.4 cm, and downstream gas-phase reaction distances of 7.3, 12.4, 17.5, and 22.5 cm were evaluated. Measurements of carbon monoxide, unburned hydrocarbons, nitrogen oxides, and pressure drop were made. The catalytic-reactor pressure drop was less than 1 percent of the upstream total pressure for all test configurations and test conditions. Nitrogen oxides and unburned hydrocarbons emissions were less than 0.25 g NO2/kg fuel and 0.6 g HC/kg fuel, respectively. The minimum operating temperature (defined as the adiabatic combustion temperature required to obtain carbon monoxide emissions below a reference level of 13.6 g CO/kg fuel) ranged from 1230 K to 1500 K for the various conditions and configurations tested. The minimum operating temperature decreased with increasing total (catalytic-reactor-plus-downstream-gas-phase-reactor-zone) residence time but was independent of the relative times spent in each region when the catalytic-reactor residence time was greater than or equal to 1.4 ms.

EAST FACE OF REACTOR BASE. COMING TOWARD CAMERA IS EXCAVATION ...

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

EAST FACE OF REACTOR BASE. COMING TOWARD CAMERA IS EXCAVATION FOR MTR CANAL. CAISSONS FLANK EACH SIDE. COUNTERFORT (SUPPORT PERPENDICULAR TO WHAT WILL BE THE LONG WALL OF THE CANAL) RESTS ATOP LEFT CAISSON. IN LOWER PART OF VIEW, DRILLERS PREPARE TRENCHES FOR SUPPORT BEAMS THAT WILL LIE BENEATH CANAL FLOOR. INL NEGATIVE NO. 739. Unknown Photographer, 10/6/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. ON GROUND FLOOR. THE 60TON ETR REACTOR VESSEL ...

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

ETR, TRA-642. ON GROUND FLOOR. THE 60-TON ETR REACTOR VESSEL IS DROPPED INTO PLACE OVER PIT. KAISER USED TWO MULTI-BLOCK DRUM PULLEYS WITH A COMBINED CAPACITY OF 100 TONS AND A 100-TON DRUM HOIST. THE VESSEL WAS 35 1/2 FEET LONG. INL NEGATIVE NO. 56-4149. R.G. Larsen, Photographer, 12/18/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

LOFT. Reactor arrives at containment building (TAN650), now being pushed ...

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

LOFT. Reactor arrives at containment building (TAN-650), now being pushed by locomotive. Camera facing northerly. Note "Hello Dolly" and "PWR MTA No. 1" (pressurized water reactor mobile test assembly) signs. Date: 1973. INEEL negative no. 73-3710 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. SOUTH SIDE ...

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

ETR BUILDING, TRA-642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. SOUTH SIDE OF ETR REACTOR, CAMERA FACING NORTH. CABINET CONTAINING "NUCLEAR INSTRUMENT SYSTEMS" IS RESTRICTED. INL NEGATIVE NO. HD46-18-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

FAST CHOPPER BUILDING, TRA665. CONTEXTUAL VIEW: CHOPPER BUILDING IN CENTER. ...

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

FAST CHOPPER BUILDING, TRA-665. CONTEXTUAL VIEW: CHOPPER BUILDING IN CENTER. MTR REACTOR SERVICES BUILDING,TRA-635, TO LEFT; MTR BUILDING TO RIGHT. CAMERA FACING WEST. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Comprehensive Nuclear-Test-Ban Treaty: Background and Current Developments

DTIC Science & Technology

2009-11-23

there are no plans to do so.’”5 Critics expressed concern about the implications of these policies for testing and new weapons. A statement by...opportunity to design and build new nuclear weapons, and abandon a ten-year-old moratorium on nuclear weapons testing.”6 Another critic felt that increased...cannot diagnose or remedy a problem in a warhead critical the U.S. nuclear deterrent without conducting a nuclear test.10 Similarly, a Statement of

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

SPERT I DESTRUCTIVE TEST PROGRAM SAFETY ANALYSIS REPORT

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

Spano, A.H.; Miller, R.W.

1962-06-15

The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor periodmore » range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)« less

Laboratory testing of extravascular body fluids in Croatia: a survey of the Working group for extravascular body fluids of the Croatian Society of Medical Biochemistry and Laboratory Medicine

PubMed Central

Kopcinovic, Lara Milevoj; Vogrinc, Zeljka; Kocijan, Irena; Culej, Jelena; Aralica, Merica; Jokic, Anja; Antoncic, Dragana; Bozovic, Marija

2016-01-01

Introduction We hypothesized that extravascular body fluid (EBF) analysis in Croatia is not harmonized and aimed to investigate preanalytical, analytical and postanalytical procedures used in EBF analysis in order to identify key aspects that should be addressed in future harmonization attempts. Materials and methods An anonymous online survey created to explore laboratory testing of EBF was sent to secondary, tertiary and private health care Medical Biochemistry Laboratories (MBLs) in Croatia. Statements were designed to address preanalytical, analytical and postanalytical procedures of cerebrospinal, pleural, peritoneal (ascites), pericardial, seminal, synovial, amniotic fluid and sweat. Participants were asked to declare the strength of agreement with proposed statements using a Likert scale. Mean scores for corresponding separate statements divided according to health care setting were calculated and compared. Results The survey response rate was 0.64 (58 / 90). None of the participating private MBLs declared to analyse EBF. We report a mean score of 3.45 obtained for all statements evaluated. Deviations from desirable procedures were demonstrated in all EBF testing phases. Minor differences in procedures used for EBF analysis comparing secondary and tertiary health care MBLs were found. The lowest scores were obtained for statements regarding quality control procedures in EBF analysis, participation in proficiency testing programmes and provision of interpretative comments on EBF’s test reports. Conclusions Although good laboratory EBF practice is present in Croatia, procedures for EBF analysis should be further harmonized to improve the quality of EBF testing and patient safety. PMID:27812307

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

PubMed

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

76 FR 57720 - Notice of Intent To Prepare an Environmental Impact Statement/Overseas Environmental Impact...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-16

...The Department of the Navy published a document in the Federal Register (76 FR 174) on September 8, 2011, concerning public scoping meetings to support the development of an Environmental Impact Statement/Overseas Environmental Impact Statement for the Mariana Islands Training and Testing Study Area. The document contained an incorrect scoping date.

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

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

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

2011-09-01

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

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

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

Gougar, Hans David

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each ofmore » the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.« less

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

NASA Astrophysics Data System (ADS)

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

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2014 CFR

2014-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2012 CFR

2012-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2010 CFR

2010-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2013 CFR

2013-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

Features of postfailure fuel behavior in transient overpower and transient undercooled/overpower tests in the transient reactor test facility

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

Doerner, R.C.; Bauer, T.H.; Morman, J.A.

Prototypic oxide fuel was subjected to simulated, fast reactor severe accident conditions in a series of in-pile tests in the Transient Reactor Test Facility reactor. Seven experiments were performed on fresh and previously irradiated oxide fuel pins under transient overpower and transient undercooled. overpower accident conditions. For each of the tests, fuel motions were observed by the hodoscope. Hodoscope data are correlated with coolant flow, pressure, and temperature data recorded by the loop instrumentation. Data were analyzed from the onset of initial failure to a final mass distribution at the end of the test. In this paper results of thesemore » analyses are compared to pre- and posttest accident calculations and to posttest metallographic accident calculations and to posttest metallographic examinations and computed tomographic reconstructions from neutron radiographs.« less

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

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

Not Available

1979-06-01

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

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

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

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

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

Cao, Yan; Gohar, Yousry

2015-11-01

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

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

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« less

MTR, TRA603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTYMETER CHOPPER ...

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

MTR, TRA-603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTY-METER CHOPPER HOUSE. COFFIN TURNING ROLLS. REMOVABLE PANEL OVER CANAL ON EAST SIDE. NEW PLUG STORAGE ACCESS. DOOR SCHEDULE INDICATES STEEL (FOR VAULT), WIRE MESH, AND HOLLOW METAL TYPES. STORAGE AND ISSUE ROOM. SAFETY SHOWERS. DOORWAY TO WING, TRA-604. BLAW-KNOX 3150-803-2, 7/1950. INL INDEX NO. 531-0603-00-098-100561, REV. 10. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

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

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

1983-06-01

During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

REACTOR SERVICE BUILDING, TRA635, INTERIOR. CAMERA FACES NORTHWEST TOWARDS INTERIOR ...

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

REACTOR SERVICE BUILDING, TRA-635, INTERIOR. CAMERA FACES NORTHWEST TOWARDS INTERIOR WALL ENCLOSING STORAGE AND OFFICE SPACE ALONG THE WEST SIDE. AT RIGHT EDGE IS DOOR TO MTR BUILDING. FROM RIGHT TO LEFT, SPACE WAS PLANNED FOR A LOCKER ROOM, MTR ISSUE ROOM, AND STORAGE AREAS AND RELATED OFFICES. NOTE SECOND "MEZZANINE" FLOOR ABOVE. INL NEGATIVE NO. 10227. Unknown Photographer, 3/23/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

NASA-EPA automotive thermal reactor technology program

NASA Technical Reports Server (NTRS)

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

1972-01-01

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

75 FR 34219 - Revision of Fee Schedules; Fee Recovery for FY 2010

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-16

....8 $6.3 $7.5 Spent Fuel Storage/Reactor Decommissioning..... -- -- 2.7 0.2 0.2 Test and Research... 2009 fee is also shown for comparative purposes. Table V--Rebaselined Annual Fees FY2009 Annual FY 2010... Decommissioning Test and Research Reactors (Non-power 87,600 81,700 Reactors) High Enriched Uranium Fuel Facility...

ETR, TRA642. ON GROUND FLOOR. WITH OUTER THERMAL RING IN ...

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

ETR, TRA-642. ON GROUND FLOOR. WITH OUTER THERMAL RING IN PLACE AND CONDUIT PRESERVED, HIGH-DENSITY CONCRETE IS PLACED BETWEEN THE THERMAL RING AND THE OUTER REACTOR FORM. INL NEGATIVE NO. 56-2400. Jack L. Anderson, Photographer, 6/10/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

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

A K Wertsching

2012-09-01

As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Fluxmore » Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to provide additional confidence with the results. The actual corrosion rates of UMo fuel is very likely to be lower than assumed within this report which can be confirmed with additional testing.« less

HISTORICAL AMERICAN ENGINEERING RECORD - IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY, TEST AREA NORTH, HAER NO. ID-33-E

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

Susan Stacy; Hollie K. Gilbert

2005-02-01

Test Area North (TAN) was a site of the Aircraft Nuclear Propulsion (ANP) Project of the U.S. Air Force and the Atomic Energy Commission. Its Cold War mission was to develop a turbojet bomber propelled by nuclear power. The project was part of an arms race. Test activities took place in five areas at TAN. The Assembly & Maintenance area was a shop and hot cell complex. Nuclear tests ran at the Initial Engine Test area. Low-power test reactors operated at a third cluster. The fourth area was for Administration. A Flight Engine Test facility (hangar) was built to housemore » the anticipated nuclear-powered aircraft. Experiments between 1955-1961 proved that a nuclear reactor could power a jet engine, but President John F. Kennedy canceled the project in March 1961. ANP facilities were adapted for new reactor projects, the most important of which were Loss of Fluid Tests (LOFT), part of an international safety program for commercial power reactors. Other projects included NASA's Systems for Nuclear Auxiliary Power and storage of Three Mile Island meltdown debris. National missions for TAN in reactor research and safety research have expired; demolition of historic TAN buildings is underway.« less

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

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

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

2004-02-04

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

Effects of glucose on the performance of enhanced biological phosphorus removal activated sludge enriched with acetate.

PubMed

Gebremariam, Seyoum Yami; Beutel, Marc W; Christian, David; Hess, Thomas F

2012-10-01

The effects of glucose on enhanced biological phosphorus removal (EBPR) activated sludge enriched with acetate was investigated using sequencing batch reactors. A glucose/acetate mixture was serially added to the test reactor in ratios of 25/75%, 50/50%, and 75/25% and the EBPR activity was compared to the control reactor fed with 100% acetate. P removal increased at a statistically significant level to a near-complete in the test reactor when the mixture increased to 50/50%. However, EBPR deteriorated when the glucose/acetate mixture increased to 75/25% in the test reactor and when the control reactor abruptly switched to 100% glucose. These results, in contrast to the EBPR conventional wisdom, suggest that the addition of glucose at moderate levels in wastewaters does not impede and may enhance EBPR, and that glucose waste products should be explored as an economical sustainable alternative when COD enhancement of EBPR is needed. Copyright © 2012 Elsevier Ltd. All rights reserved.

Flat-plate collector research area: Silicon material task

NASA Technical Reports Server (NTRS)

Lutwack, R.

1982-01-01

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

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

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

M. Chen; CM Regan; D. Noe

2006-01-09

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas releasemore » and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.« less

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

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

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the Highmore » Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.« less

Advanced Thermal Simulator Testing: Thermal Analysis and Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Reid, Robert; Adams, Mike; Davis, Joe

2008-01-01

Work at the NASA Marshall Space Flight Center seeks to develop high fidelity, electrically heated thermal simulators that represent fuel elements in a nuclear reactor design to support non-nuclear testing applicable to the development of a space nuclear power or propulsion system. Comparison between the fuel pins and thermal simulators is made at the outer fuel clad surface, which corresponds to the outer sheath surface in the thermal simulator. The thermal simulators that are currently being tested correspond to a SNAP derivative reactor design that could be applied for Lunar surface power. These simulators are designed to meet the geometric and power requirements of a proposed surface power reactor design, accommodate testing of various axial power profiles, and incorporate imbedded instrumentation. This paper reports the results of thermal simulator analysis and testing in a bare element configuration, which does not incorporate active heat removal, and testing in a water-cooled calorimeter designed to mimic the heat removal that would be experienced in a reactor core.

The STARD statement for reporting diagnostic accuracy studies: application to the history and physical examination.

PubMed

Simel, David L; Rennie, Drummond; Bossuyt, Patrick M M

2008-06-01

The Standards for Reporting of Diagnostic Accuracy (STARD) statement provided guidelines for investigators conducting diagnostic accuracy studies. We reviewed each item in the statement for its applicability to clinical examination diagnostic accuracy research, viewing each discrete aspect of the history and physical examination as a diagnostic test. Nonsystematic review of the STARD statement. Two former STARD Group participants and 1 editor of a journal series on clinical examination research reviewed each STARD item. Suggested interpretations and comments were shared to develop consensus. The STARD Statement applies generally well to clinical examination diagnostic accuracy studies. Three items are the most important for clinical examination diagnostic accuracy studies, and investigators should pay particular attention to their requirements: describe carefully the patient recruitment process, describe participant sampling and address if patients were from a consecutive series, and describe whether the clinicians were masked to the reference standard tests and whether the interpretation of the reference standard test was masked to the clinical examination components or overall clinical impression. The consideration of these and the other STARD items in clinical examination diagnostic research studies would improve the quality of investigations and strengthen conclusions reached by practicing clinicians. The STARD statement provides a very useful framework for diagnostic accuracy studies. The group correctly anticipated that there would be nuances applicable to studies of the clinical examination. We offer guidance that should enhance their usefulness to investigators embarking on original studies of a patient's history and physical examination.

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

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

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

2D Thermal Hydraulic Analysis and Benchmark in Support of HFIR LEU Conversion using COMSOL

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

Freels, James D; Bodey, Isaac T; Lowe, Kirk T

2010-09-01

The research documented herein was funded by a research contract between the Research Reactors Division (RRD) of Oak Ridge National Laboratory (ORNL) and the University of Tennessee, Knoxville (UTK) Mechanical, Aerospace and Biomedical Engineering Department (MABE). The research was governed by a statement of work (SOW) which clearly defines nine specific tasks. This report is outlined to follow and document the results of each of these nine specific tasks. The primary goal of this phase of the research is to demonstrate, through verification and validation methods, that COMSOL is a viable simulation tool for thermal-hydraulic modeling of the High Fluxmore » Isotope Reactor (HFIR) core. A secondary goal of this two-dimensional phase of the research is to establish methodology and data base libraries that are also needed in the full three-dimensional COMSOL simulation to follow. COMSOL version 3.5a was used for all of the models presented throughout this report.« less

NEUTRONIC REACTOR SHIELDING

DOEpatents

Borst, L.B.

1961-07-11

A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2010 CFR

2010-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

Reactor performance and microbial community dynamics during anaerobic co-digestion of municipal wastewater sludge with restaurant grease waste at steady state and overloading stages.

PubMed

Razaviarani, Vahid; Buchanan, Ian D

2014-11-01

Linkage between reactor performance and microbial community dynamics was investigated during mesophilic anaerobic co-digestion of restaurant grease waste (GTW) with municipal wastewater sludge (MWS) using 10L completely mixed reactors and a 20day SRT. Test reactors received a mixture of GTW and MWS while control reactors received only MWS. Addition of GTW to the test reactors enhanced the biogas production and methane yield by up to 65% and 120%, respectively. Pyrosequencing revealed that Methanosaeta and Methanomicrobium were the dominant acetoclastic and hydrogenotrophic methanogen genera, respectively, during stable reactor operation. The number of Methanosarcina and Methanomicrobium sequences increased and that of Methanosaeta declined when the proportion of GTW in the feed was increased to cause an overload condition. Under this overload condition, the pH, alkalinity and methane production decreased and VFA concentrations increased dramatically. Candidatus cloacamonas, affiliated within phylum Spirochaetes, were the dominant bacterial genus at all reactor loadings. Copyright © 2014 Elsevier Ltd. All rights reserved.

ETRCF, TRA654, INTERIOR. CAMERA IS ON MAIN FLOOR. NOTE CRANE ...

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

ETR-CF, TRA-654, INTERIOR. CAMERA IS ON MAIN FLOOR. NOTE CRANE HOOKS. ELECTRICAL EQUIPMENT IS PART OF PAST EXPERIMENT. DOOR AT LEFT EDGE OF VIEW LEADS TO REACTOR SERVICE BUILDING, TRA-635. INL NEGATIVE NO. HD24-1-2. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

OAST Space Theme Workshop. Volume 2: Theme summary. 4: Solar system exploration (no. 10). A: Statement of theme: B. 26 April 1976 Presentation. C. Summary. D. Initiative actions (form 5)

NASA Technical Reports Server (NTRS)

1976-01-01

Major strategies for exploring the solar system focus on the return of information and the return of matter. Both the planetary exploration facility, and an orbiting automated space station, and the sample return and exploration facility have similar requirements. The single most essential need to enable intensive study of the outer solar system is nuclear propulsion and power capability. New initiatives in 1978 related to the reactor, data and sample acquisition and return, navigation, and environmental protection are examined.

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

Cappiello, M.; Hobbins, R.; Penny, K.

As part of the Department of Energy Advanced Fuel Cycle program, a series of fuels development irradiation tests have been performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. These tests are providing excellent data for advanced fuels development. The program is focused on the transmutation of higher actinides which best can be accomplished in a sodium-cooled fast reactor. Because a fast test reactor is no longer available in the US, a special test vehicle is used to achieve near-prototypic fast reactor conditions (neutron spectra and temperature) for use in ATR (a water-cooled thermal reactor). As partmore » of the testing program, there were many successful tests of advanced fuels including metals and ceramics. Recently however, there have been three experimental campaigns using metal fuels that experienced failure during irradiation. At the request of the program, an independent review committee was convened to review the post-test analyses performed by the fuels development team, to assess the conclusions of the team for the cause of the failures, to assess the adequacy and completeness of the analyses, to identify issues that were missed, and to make recommendations for improvements in the design and operation of future tests. Although there is some difference of opinion, the review committee largely agreed with the conclusions of the fuel development team regarding the cause of the failures. For the most part, the analyses that support the conclusions are sufficient.« less

The development of an instrument to assess chemistry perceptions

NASA Astrophysics Data System (ADS)

Wells, Raymond R.

The instrument, developed in this study, attempted to correct the deficiencies of previous instruments. Statements of belief and opinion can be validly included under the construct of chemistry perceptions. Further, statements that might be better characterized as science attitudes, math attitudes, or attitudes toward a specific course or program were not included. Eliminating statements of math anxiety and test anxiety insured that responses to statements of anxiety were perceptions of anxiety solely related to chemistry. The results of the expert judges' responses to the Validation of Proposed Perception Statements forms were detailed to establish construct and content validity. The nature of Likert scale construction and calculation of internal consistency also supported the validity of the instrument. A pilot Chemistry Perception Questionnaire (CPQ) was then constructed based on agreement of the appropriate subscale and mean importance of the perception statements. The pilot CPQ results were subjected to an item analysis based on three sets of statistics: the frequency of each response and the percentage of respondents making each response for each perception statement, the mean and standard deviations for each item, and the item discrimination index which correlated the item scores with the subscale scores. With no zero or negative correlations to the subscale scores, it was not necessary to replace any of the perception statements contained in the pilot instrument. Therefore, the piloted Chemistry Perception Questionnaire became the final instrument. Factor analysis confirmed the multidimensionality of the instrument. The instrument was administered twice with a separation interval of approximately one month in order to perform a test-retest reliability analysis. One hundred and forty-one pairs were matched and results detailed. The correlation between forms, for the total instrument, was 0.9342. The mean coefficient alpha, for the total instrument, was 0.9495. With test-retest correlations and alphas exceeding 0.70 for all seven subscales and the total instrument, it was determined that the Chemistry Perception Questionnaire instrument achieved reasonably high reliability estimations.

Identifying Universal Linguistic Features Associated with Veracity and Deception

DTIC Science & Technology

2015-01-02

involved an eyewitness testing paradigm, in which participants from three very different language groups witnessed an actual crime and write true and...statements will include significantly more indicators of deception and significantly less indicators of veracity across all languages tested ...examine possible differential carry-over effects if participants wrote both true and false statements in a within-subjects design . The findings

Easy and Informative: Using Confidence-Weighted True-False Items for Knowledge Tests in Psychology Courses

ERIC Educational Resources Information Center

Dutke, Stephan; Barenberg, Jonathan

2015-01-01

We introduce a specific type of item for knowledge tests, confidence-weighted true-false (CTF) items, and review experiences of its application in psychology courses. A CTF item is a statement about the learning content to which students respond whether the statement is true or false, and they rate their confidence level. Previous studies using…

One-Year Test-Retest Reliability of the Inventory of Statements about Self-Injury (ISAS)

ERIC Educational Resources Information Center

Glenn, Catherine R.; Klonsky, E. David

2011-01-01

Nonsuicidal self-injury (NSSI) is a growing public health problem among adolescents and young adults. The Inventory of Statements About Self-Injury (ISAS) is a self-report measure designed to assess NSSI behaviors and functions. The current study examines the one-year test-retest reliability of the ISAS in a sample of young adult self-injurers.…

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

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

MH Lane

2006-02-15

This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

In-reactor oxidation of zircaloy-4 under low water vapor pressures

NASA Astrophysics Data System (ADS)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor

PubMed Central

Zhang, Shubin; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance. PMID:29121067

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor.

PubMed

Zhang, Shubin; Zhang, Yufeng; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance.

A Review of Gas-Cooled Reactor Concepts for SDI Applications

DTIC Science & Technology

1989-08-01

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

Responses to diagnostic tests for bovine tuberculosis in dairy and non-dairy cattle naturally exposed to Mycobacterium bovis in Great Britain.

PubMed

Downs, S H; Broughan, J M; Goodchild, A V; Upton, P A; Durr, P A

2016-10-01

Field surveillance of British cattle using the single intradermal comparative cervical tuberculin (SICCT) test shows a higher incidence rate of bovine tuberculosis (bTB) in dairy compared to beef herds, but a lower probability of post-mortem examination confirmed (PMC) Mycobacterium bovis infection in dairy herds. A cross-sectional study was conducted to compare animal level differences in bTB detection between dairy and non-dairy cattle in Great Britain. During the period from 2002 to 2005, 200 (41% dairy) reactors in the SICCT test (standard interpretation) were randomly selected, and 200 in-contact cattle (43% dairy) were purposively selected from bTB-infected herds. Interferon (IFN)-γ responses in blood to bovine and avian purified protein derivative (PPD), and early secretory antigen target 6 kDa and culture filtrate protein 10 (ESAT-6/CFP10), were measured. The post-mortem examination included gross pathological examination, mycobacterial culture and histopathology. The proportions of cattle positive to ESAT6/CFP10 were 26% (95% confidence interval, CI, 15-39%) in dairy reactors and 62% (95% CI 51-72%) in non-dairy reactors (P <0.001). PMC risk was 34% (95% CI 24-45%) in dairy reactors and 69% (95% CI 60-78%) in non-dairy reactors (P <0.001). The odds ratio for PMC risk in dairy reactors compared to non-dairy reactors, after controlling for bTB prevalence, herd size and SICCT test response, was 0.27 (95% CI 0.14-0.53; P <0.001). In surveillance data, adjusted animal level PMC risks were lower for dairy reactors than for beef reactors aged >2 years (P <0.001). Crown Copyright © 2016. Published by Elsevier Ltd. All rights reserved.

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

Paradoxical effects of alcohol information on alcohol outcome expectancies.

PubMed

Krank, Marvin D; Ames, Susan L; Grenard, Jerry L; Schoenfeld, Tara; Stacy, Alan W

2010-07-01

Cognitive associations with alcohol predict both current and future use in youth and young adults. Much cognitive and social cognitive research suggests that exposure to information may have unconscious influences on thinking and behavior. The present study assessed the impact of information statements on the accessibility of alcohol outcome expectancies. The 2 studies reported here investigated the effects of exposure to alcohol statements typical of informational approaches to prevention on the accessibility of alcohol outcome expectancies. High school and university students were presented with information statements about the effects of alcohol and other commercial products. The alcohol statements were taken from expectancy questionnaires. Some of these statements were presented as facts and others as myths. The retention of detailed information about these statements was manipulated by (i) divided attention versus focused attention or (ii) immediate versus delayed testing. Accessibility of personal alcohol outcome expectancies was subsequently measured using an open-ended question about the expected effects of alcohol. Participants reported more alcohol outcomes seen during the information task as personal expectations about the effects of alcohol use than similar unseen items. Paradoxically, myth statements were also more likely to be reported as expectancies than unseen items in all conditions. Additionally, myth statements were generated less often than fact statements only under the condition of immediate testing with strong content processing instructions. These observations are consistent with findings from cognitive research where familiarity in the absence of explicit memory can have an unconscious influence on performance. In particular, the exposure to these items in an informational format increases accessibility of the seen items even when the participants were told that they were myths. The findings have implications for the development of effective prevention materials.

Paradoxical Effects of Alcohol Information on Alcohol Outcome Expectancies

PubMed Central

Krank, Marvin D.; Ames, Susan L.; Grenard, Jerry L.; Schoenfeld, Tara; Stacy, Alan W.

2014-01-01

Background Cognitive associations with alcohol predict both current and future use in youth and young adults. Much cognitive and social cognitive research suggests that exposure to information may have unconscious influences on thinking and behavior. The present study assessed the impact of information statements on the accessibility of alcohol outcome expectancies. Methods The 2 studies reported here investigated the effects of exposure to alcohol statements typical of informational approaches to prevention on the accessibility of alcohol outcome expectancies. High school and university students were presented with information statements about the effects of alcohol and other commercial products. The alcohol statements were taken from expectancy questionnaires. Some of these statements were presented as facts and others as myths. The retention of detailed information about these statements was manipulated by (i) divided attention versus focused attention or (ii) immediate versus delayed testing. Accessibility of personal alcohol outcome expectancies was subsequently measured using an open-ended question about the expected effects of alcohol. Results Participants reported more alcohol outcomes seen during the information task as personal expectations about the effects of alcohol use than similar unseen items. Paradoxically, myth statements were also more likely to be reported as expectancies than unseen items in all conditions. Additionally, myth statements were generated less often than fact statements only under the condition of immediate testing with strong content processing instructions. Conclusions These observations are consistent with findings from cognitive research where familiarity in the absence of explicit memory can have an unconscious influence on performance. In particular, the exposure to these items in an informational format increases accessibility of the seen items even when the participants were told that they were myths. The findings have implications for the development of effective prevention materials. PMID:20477773

Understanding of Speaker Certainty and False-Belief Reasoning: A Comparison of Japanese and German Preschoolers

ERIC Educational Resources Information Center

Matsui, Tomoko; Rakoczy, Hannes; Miura, Yui; Tomasello, Michael

2009-01-01

It has been repeatedly shown that when asked to identify a protagonist's false belief on the basis of his false statement, English-speaking 3-year-olds dismiss the statement and fail to attribute to him a false belief. In the present studies, we tested 3-year-old Japanese children in a similar task, using false statements accompanied by…

Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

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

Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.

2013-07-01

Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away frommore » reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)« less

Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

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

James E. O'Brien; Piyush Sabharwall; SuJong Yoon

2001-11-01

Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water.more » Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed« less

78 FR 48501 - Agency Information Collection Activities: Proposed Collection; Comment Request

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-08

... storage installations, decommissioned power reactors, power reactors under construction, research and test reactors, agreement states, non-agreement states, as well as departments of health, medical centers, steel...

MTR,TRA603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED ...

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

MTR,TRA-603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED CUBICLES WERE IDENTIFIED BY SPONSORING LABORATORY AND ITS TEST HOLE NUMBER IN THE REACTOR, IE, "KAPL HB-1" SIGNIFIED KNOLLS ATOMIC POWER LABORATORY, HORIZONTAL BEAM NO. 1. "WAPD" WAS WESTINGHOUSE ATOMIC POWER DIVISION. CATCH TANKS AND SAMPLE STATIONS FOR TEST LOOPS WERE ASSOCIATED WITH THESE CUBICLES. NOTE DESKS, STORAGE CABINETS, SWITCH GEAR, INSTRUMENT PANELS. PHILLIPS PETROLEUM COMPANY MTR-E-5205, 4/1963. INL INDEX NO. 531-0603-00-706-009757, REV. 5. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

NASA Technical Reports Server (NTRS)

Springborn, R. H.

1971-01-01

Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

Medical records for animals used in research, teaching, and testing: public statement from the American College of Laboratory Animal Medicine.

PubMed

Field, Karl; Bailey, Michele; Foresman, Larry L; Harris, Robert L; Motzel, Sherri L; Rockar, Richard A; Ruble, Gaye; Suckow, Mark A

2007-01-01

Medical records are considered to be a key element of a program of adequate veterinary care for animals used in research, teaching, and testing. However, prior to the release of the public statement on medical records by the American College of Laboratory Animal Medicine (ACLAM), the guidance that was available on the form and content of medical records used for the research setting was not consistent and, in some cases, was considered to be too rigid. To address this concern, ACLAM convened an ad hoc Medical Records Committee and charged the Committee with the task of developing a medical record guideline that was based on both professional judgment and performance standards. The Committee provided ACLAM with a guidance document titled Public Statements: Medical Records for Animals Used in Research, Teaching, and Testing, which was approved by ACLAM in late 2004. The ACLAM public statement on medical records provides guidance on the definition and content of medical records, and clearly identifies the Attending Veterinarian as the individual who is charged with authority and responsibility for oversight of the institution's medical records program. The document offers latitude to institutions in the precise form and process used for medical records but identifies typical information to be included in such records. As a result, the ACLAM public statement on medical records provides practical yet flexible guidelines to assure that documentation of animal health is performed in research, teaching, and testing situations.

Enhanced In-Pile Instrumentation at the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

2012-08-01

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

ETR HEAT EXCHANGER BUILDING, TRA644. FLOOR PLAN AND SECTIONS. PUMP ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. FLOOR PLAN AND SECTIONS. PUMP CUBICLES WITH PUMP MOTORS OUTSIDE CUBICLES. HEAT EXCHANGER EQUIPMENT. COOLANT PIPE TUNNEL ENTERS FROM REACTOR BUILDING. KAISER ETR-5582-MTR-644-A-3, 2/1956. INL INDEX NO. 532-0644-00-486-101294, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

High-Temperature Gas-Cooled Test Reactor Point Design

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

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

2016-04-01

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

ETR, TRA642. NORTHSOUTH SECTION, LOOKING WEST. STEELFRAME ROOF, CRANE RAIL, ...

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

ETR, TRA-642. NORTH-SOUTH SECTION, LOOKING WEST. STEEL-FRAME ROOF, CRANE RAIL, AND CRANES. COOLANT PIPE TUNNEL LEADING TO REACTOR FROM EAST. (THIS WAS A PRELIMINARY CONCEPT DRAWING.) KAISER ETR-5528-MTR-642-A-4, 11/1955. INL INDEX NO. 532-0642-00-486-100912, REV. 1. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

10 CFR 100.10 - Factors to be considered when evaluating sites.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Section 100.10 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactors § 100... include those relating both to the proposed reactor design and the characteristics peculiar to the site...

10 CFR 100.10 - Factors to be considered when evaluating sites.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Section 100.10 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactors § 100... include those relating both to the proposed reactor design and the characteristics peculiar to the site...

REACTOR FUEL ELEMENTS TESTING CONTAINER

DOEpatents

Whitham, G.K.; Smith, R.R.

1963-01-15

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Progress towards developing neutron tolerant magnetostrictive and piezoelectric transducers

NASA Astrophysics Data System (ADS)

Reinhardt, Brian; Tittmann, Bernhard; Rempe, Joy; Daw, Joshua; Kohse, Gordon; Carpenter, David; Ames, Michael; Ostrovsky, Yakov; Ramuhalli, Pradeep; Montgomery, Robert; Chien, Hualte; Wernsman, Bernard

2015-03-01

Current generation light water reactors (LWRs), sodium cooled fast reactors (SFRs), small modular reactors (SMRs), and next generation nuclear plants (NGNPs) produce harsh environments in and near the reactor core that can severely tax material performance and limit component operational life. To address this issue, several Department of Energy Office of Nuclear Energy (DOE-NE) research programs are evaluating the long duration irradiation performance of fuel and structural materials used in existing and new reactors. In order to maximize the amount of information obtained from Material Testing Reactor (MTR) irradiations, DOE is also funding development of enhanced instrumentation that will be able to obtain in-situ, real-time data on key material characteristics and properties, with unprecedented accuracy and resolution. Such data are required to validate new multi-scale, multi-physics modeling tools under development as part of a science-based, engineering driven approach to reactor development. It is not feasible to obtain high resolution/microscale data with the current state of instrumentation technology. However, ultrasound-based sensors offer the ability to obtain such data if it is demonstrated that these sensors and their associated transducers are resistant to high neutron flux, high gamma radiation, and high temperature. To address this need, the Advanced Test Reactor National Scientific User Facility (ATR-NSUF) is funding an irradiation, led by PSU, at the Massachusetts Institute of Technology Research Reactor to test the survivability of ultrasound transducers. As part of this effort, PSU and collaborators have designed, fabricated, and provided piezoelectric and magnetostrictive transducers that are optimized to perform in harsh, high flux, environments. Four piezoelectric transducers were fabricated with either aluminum nitride, zinc oxide, or bismuth titanate as the active element that were coupled to either Kovar or aluminum waveguides and two magnetostrictive transducers were fabricated with Remendur or Galfenol as the active elements. Pulse-echo ultrasonic measurements of these transducers are made in-situ. This paper will present an overview of the test design including selection criteria for candidate materials and optimization of test assembly parameters, data obtained from both out-of-pile and in-pile testing at elevated temperatures, and an assessment based on initial data of the expected performance of ultrasonic devices in irradiation conditions.

Review of Nuclear Thermal Propulsion Ground Test Options

NASA Technical Reports Server (NTRS)

Coote, David J.; Power, Kevin P.; Gerrish, Harold P.; Doughty, Glen

2015-01-01

High efficiency rocket propulsion systems are essential for humanity to venture beyond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rockets with relatively high thrust and twice the efficiency of highest performing chemical propellant engines. NTP utilizes the coolant of a nuclear reactor to produce propulsive thrust. An NTP engine produces thrust by flowing hydrogen through a nuclear reactor to cool the reactor, heating the hydrogen and expelling it through a rocket nozzle. The hot gaseous hydrogen is nominally expected to be free of radioactive byproducts from the nuclear reactor; however, it has the potential to be contaminated due to off-nominal engine reactor performance. NTP ground testing is more difficult than chemical engine testing since current environmental regulations do not allow/permit open air testing of NTP as was done in the 1960's and 1970's for the Rover/NERVA program. A new and innovative approach to rocket engine ground test is required to mitigate the unique health and safety risks associated with the potential entrainment of radioactive waste from the NTP engine reactor core into the engine exhaust. Several studies have been conducted since the ROVER/NERVA program in the 1970's investigating NTP engine ground test options to understand the technical feasibility, identify technical challenges and associated risks and provide rough order of magnitude cost estimates for facility development and test operations. The options can be divided into two distinct schemes; (1) real-time filtering of the engine exhaust and its release to the environment or (2) capture and storage of engine exhaust for subsequent processing.

Impact of Substratum Surface on Microbial Community Structure and Treatment Performance in Biological Aerated Filters

PubMed Central

Kim, Lavane; Pagaling, Eulyn; Zuo, Yi Y.

2014-01-01

The impact of substratum surface property change on biofilm community structure was investigated using laboratory biological aerated filter (BAF) reactors and molecular microbial community analysis. Two substratum surfaces that differed in surface properties were created via surface coating and used to develop biofilms in test (modified surface) and control (original surface) BAF reactors. Microbial community analysis by 16S rRNA gene-based PCR-denaturing gradient gel electrophoresis (DGGE) showed that the surface property change consistently resulted in distinct profiles of microbial populations during replicate reactor start-ups. Pyrosequencing of the bar-coded 16S rRNA gene amplicons surveyed more than 90% of the microbial diversity in the microbial communities and identified 72 unique bacterial species within 19 bacterial orders. Among the 19 orders of bacteria detected, Burkholderiales and Rhodocyclales of the Betaproteobacteria class were numerically dominant and accounted for 90.5 to 97.4% of the sequence reads, and their relative abundances in the test and control BAF reactors were different in consistent patterns during the two reactor start-ups. Three of the five dominant bacterial species also showed consistent relative abundance changes between the test and control BAF reactors. The different biofilm microbial communities led to different treatment efficiencies, with consistently higher total organic carbon (TOC) removal in the test reactor than in the control reactor. Further understanding of how surface properties affect biofilm microbial communities and functional performance would enable the rational design of new generations of substrata for the improvement of biofilm-based biological treatment processes. PMID:24141134

Design test request No. 1263 K Reactor graphite key and VSR channel sleeve test

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

Kempf, F.J.

1964-12-10

The objectives of this test were: (1) Determine the coefficient of friction between two adjacent layers of K Reactor graphite at room temperature. (2) Determine the average load required to cause failure of an unirradiated K Reactor side reflector bar, when subjected to tensile loading applied through the reflector keys. (3) Determine the average load at failure and the average deflection at failure of a single VSR channel key when loaded in keyways with clearances equal to those used in original stack construction. (4) Determine the average load and deflection required to break the four K Reactor VSR keys whenmore » loaded simultaneously in both `3-layer` and `7-layer` mockups. Also determine the mode of key failure; i.e., shear, flexure or combined compression and bending. Following these key rupture tests, determine the strength and deflection characteristics of the proposed K Reactor VSR channel sleeve when loaded in a manner identical to that used to fracture the keys. (5) Determine the average load and deflection at failure of both the proposed K Reactor VSR channel sleeves and the proposed C Reactor sleeves when subjected to crushing loads. (6) Determine the extent of damage to the proposed K Reactor VSR channel sleeve when subjected to the following vertical rod loading conditions. (a) Full rod drop in a channel mockup which has been misaligned 2 1/2 inches. (b) Full rod drop in a channel which has been misaligned an amount equal to the maximum flexibility of a `universal` VSR.« less

Recent Declines in Infant Mortality in the United States, 2005-2011

MedlinePlus

... 37 completed weeks of gestation. Data source and methods Data presented in this report were based on ... Text statements were tested for statistical significance using methods described elsewhere ( 3 ), and a statement that a ...

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

DTIC Science & Technology

2008-11-03

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

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

DTIC Science & Technology

2008-10-02

8 indigenous Indian power reactors ! Fast Breeder test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction ! Enrichment... breeder reactors could be viewed as providing a significant nonproliferation benefit because the materials produced by these plants are a few steps closer...to potential use in a bomb. In addition, safeguards on enrichment, reprocessing plants, and breeder reactors would support the 2002 U.S. National

SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

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

IJ van Rooyen; WR Lloyd; TL Trowbridge

2013-09-01

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designsmore » being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points and clad configurations. The 2-ply sleeve samples show a higher bend momentum compared to those of the 1-ply sleeve samples. This is applicable to both the hybrid mock-up and bare SiC-CMC sleeve samples. Comparatively both the 1- and 2-ply hybrid mock-up samples showed a higher bend stiffness and strength compared with the standard Zr-4 mock-up sample. The characterization of the hybrid mock-up samples showed signs of distress and preliminary signs of fraying at the protective Zr-4 sleeve areas for the 1-ply SiC-CMC sleeve. In addition, the microstructure of the SiC matrix near the cracks at the region of highest compressive bending strain shows significant cracking and flaking. The 2-ply SiC-CMC sleeve samples showed a more bonded, cohesive SiC matrix structure. This cracking and fraying causes concern for increased fretting during the actual use of the design. Tomography was proven as a successful tool to identify open porosity during pre-test characterization. Although there is currently insufficient data to make conclusive statements regarding the overall merit of the hybrid cladding design, preliminary characterization of this novel design has been demonstrated.« less

Gas phase oxidation downstream of a catalytic combustor

NASA Technical Reports Server (NTRS)

Tien, J. S.; Anderson, D. N.

1979-01-01

Effect of the length available for gas-phase reactions downstream of the catalytic reactor on the emission of CO and unburned hydrocarbons was investigated. A premixed, prevaporized propane/air feed to a 12/cm/diameter catalytic/reactor test section was used. The catalytic reactor was made of four 2.5 cm long monolithic catalyst elements. Four water cooled gas sampling probes were located at positions between 0 and 22 cm downstream of the catalytic reactor. Measurements of unburned hydrocarbon, CO, and CO2 were made. Tests were performed with an inlet air temperature of 800 K, a reference velocity of 10 m/s, pressures of 3 and 600,000 Pa, and fuel air equivalence ratios of 0.14 to 0.24. For very lean mixtures, hydrocarbon emissions were high and CO continued to be formed downstream of the catalytic reactor. At the highest equivalence ratios tested, hydrocarbon levels were much lower and CO was oxidized to CO2 in the gas phase downstream. To achieve acceptable emissions, a downstream region several times longer than the catalytic reactor could be required.

Comparison of heavy metal toxicity in continuous flow and batch reactors

NASA Astrophysics Data System (ADS)

Sengor, S. S.; Gikas, P.; Moberly, J. G.; Peyton, B. M.; Ginn, T. R.

2009-12-01

The presence of heavy metals may significantly affect microbial growth. In many cases, small amounts of particular heavy metals may stimulate microbial growth; however, larger quantities may result in microbial growth reduction. Environmental parameters, such as growth pattern may alter the critical heavy metal concentration, above which microbial growth stimulation turns to growth inhibition. Thus, it is important to quantify the effects of heavy metals on microbial activity for understanding natural or manmade biological reactors, either in situ or ex situ. Here we compare the toxicity of Zn and Cu on Arthrobacter sp., a heavy metal tolerant microorganism, under continuous flow versus batch reactor operations. Batch and continuous growth tests of Arthrobacter sp. were carried out at various individual and combined concentrations of Zn and Cu. Biomass concentration (OD) was measured for both the batch and continuous reactors, whereas ATP, oxygen uptake rates and substrate concentrations were additionally measured for the continuous system. Results indicated that Cu was more toxic than Zn under all conditions for both systems. In batch reactors, all tested Zn concentrations up to 150 uM showed a stimulatory effect on microbial growth. However, in the case of mixed Zn and Cu exposures, the presence of Zn either eliminated (at the 50 uM level both Zn and Cu) or reduced by ~25% (at the 100 and 150 uM levels both Zn and Cu) the Cu-induced inhibition. In the continuous system, only one test involved combined Cu (40uM) and Zn (125uM) and this test showed similar results to the 40uM Cu continuous test, i.e., no reduction in inhibition. The specific ATP concentration, i.e., ATP/OD, results for the continuous reactor showed an apparent recovery for both Cu-treated populations, although neither the OD nor glucose data showed any recovery. This may reflect that the individual microorganisms that survived after the addition of heavy metals, kept maintaining the usual ATP levels, as before metal addition. The last may imply a short of adaptation by some microorganisms to the presence of heavy metals. Overall, the batch reactor tests underestimated significantly the heavy metal inhibition, as compared to the continuous flow reactors. Therefore, the results of batch reactor tests should be used with some caution when heavy metal inhibition is to be interpreted for continuous flow natural environmental systems, such as rivers or wetlands.

Noninvasive prenatal screening for fetal aneuploidy, 2016 update: a position statement of the American College of Medical Genetics and Genomics.

PubMed

Gregg, Anthony R; Skotko, Brian G; Benkendorf, Judith L; Monaghan, Kristin G; Bajaj, Komal; Best, Robert G; Klugman, Susan; Watson, Michael S

2016-10-01

This statement is designed primarily as an educational resource for clinicians to help them provide quality medical services. Adherence to this statement is completely voluntary and does not necessarily assure a successful medical outcome. This statement should not be considered inclusive of all proper procedures and tests or exclusive of other procedures and tests that are reasonably directed toward obtaining the same results. In determining the propriety of any specific procedure or test, the clinician should apply his or her own professional judgment to the specific clinical circumstances presented by the individual patient or specimen. Clinicians are encouraged to document the reasons for the use of a particular procedure or test, whether or not it is in conformance with this statement. Clinicians also are advised to take notice of the date this statement was adopted and to consider other medical and scientific information that becomes available after that date. It also would be prudent to consider whether intellectual property interests may restrict the performance of certain tests and other procedures.Noninvasive prenatal screening using cell-free DNA (NIPS) has been rapidly integrated into prenatal care since the initial American College of Medical Genetics and Genomics (ACMG) statement in 2013. New evidence strongly suggests that NIPS can replace conventional screening for Patau, Edwards, and Down syndromes across the maternal age spectrum, for a continuum of gestational age beginning at 9-10 weeks, and for patients who are not significantly obese. This statement sets forth a new framework for NIPS that is supported by information from validation and clinical utility studies. Pretest counseling for NIPS remains crucial; however, it needs to go beyond discussions of Patau, Edwards, and Down syndromes. The use of NIPS to include sex chromosome aneuploidy screening and screening for selected copy-number variants (CNVs) is becoming commonplace because there are no other screening options to identify these conditions. Providers should have a more thorough understanding of patient preferences and be able to educate about the current drawbacks of NIPS across the prenatal screening spectrum. Laboratories are encouraged to meet the needs of providers and their patients by delivering meaningful screening reports and to engage in education. With health-care-provider guidance, the patient should be able to make an educated decision about the current use of NIPS and the ramifications of a positive, negative, or no-call result.Genet Med 18 10, 1056-1065.

The STARD Statement for Reporting Diagnostic Accuracy Studies: Application to the History and Physical Examination

PubMed Central

Rennie, Drummond; Bossuyt, Patrick M. M.

2008-01-01

Summary Objective The Standards for Reporting of Diagnostic Accuracy (STARD) statement provided guidelines for investigators conducting diagnostic accuracy studies. We reviewed each item in the statement for its applicability to clinical examination diagnostic accuracy research, viewing each discrete aspect of the history and physical examination as a diagnostic test. Setting Nonsystematic review of the STARD statement. Interventions Two former STARD Group participants and 1 editor of a journal series on clinical examination research reviewed each STARD item. Suggested interpretations and comments were shared to develop consensus. Measurements and Main Results The STARD Statement applies generally well to clinical examination diagnostic accuracy studies. Three items are the most important for clinical examination diagnostic accuracy studies, and investigators should pay particular attention to their requirements: describe carefully the patient recruitment process, describe participant sampling and address if patients were from a consecutive series, and describe whether the clinicians were masked to the reference standard tests and whether the interpretation of the reference standard test was masked to the clinical examination components or overall clinical impression. The consideration of these and the other STARD items in clinical examination diagnostic research studies would improve the quality of investigations and strengthen conclusions reached by practicing clinicians. Conclusions The STARD statement provides a very useful framework for diagnostic accuracy studies. The group correctly anticipated that there would be nuances applicable to studies of the clinical examination. We offer guidance that should enhance their usefulness to investigators embarking on original studies of a patient’s history and physical examination. PMID:18347878

[Method for evaluating the competence of specialists--the validation of 360-degree-questionnaire].

PubMed

Nørgaard, Kirsten; Pedersen, Juri; Ravn, Lisbeth; Albrecht-Beste, Elisabeth; Holck, Kim; Fredløv, Maj; Møller, Lars Krag

2010-04-19

Assessment of physicians' performance focuses on the quality of their work. The aim of this study was to develop a valid, usable and acceptable multisource feedback assessment tool (MFAT) for hospital consultants. Statements were produced on consultant competencies within non-medical areas like collaboration, professionalism, communication, health promotion, academics and administration. The statements were validated by physicians and later by non-physician professionals after adjustments had been made. In a pilot test, a group of consultants was assessed using the final collection of statements of the MFAT. They received a report with their personal results and subsequently evaluated the assessment method. In total, 66 statements were developed and after validation they were reduced and reformulated to 35. Mean scores for relevance and "easy to understand" of the statements were in the range between "very high degree" and "high degree". In the pilot test, 18 consultants were assessed by themselves, by 141 other physicians and by 125 other professionals in the hospital. About two thirds greatly benefited of the assessment report and half identified areas for personal development. About a third did not want the head of their department to know the assessment results directly; however, two thirds found a potential value in discussing the results with the head. We developed an MFAT for consultants with relevant and understandable statements. A pilot test confirmed that most of the consultants gained from the assessment, but some did not like to share their results with their heads. For these specialists other methods should be used.

Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

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

K. L. Davis; D. L. Knudson; J. L. Rempe

New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status ofmore » INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.« less

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

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

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

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

Evaluation of some candidate materials for automobile thermal reactors in engine-dynamometer screening tests

NASA Technical Reports Server (NTRS)

Oldrieve, R. E.

1971-01-01

Fourteen materials were evaluated in engine screening tests on full-size thermal reactors for automobile engine pollution control systems. Cyclic test-stand engine operation provided 2 hours at 1040 C and a 20-minute air-cool to 70 C each test cycle. Each reactor material was exposed to 83 cycles in 200 hours of engine testing. On the basis of resistance to oxidation and distortion, the best materials included two ferritic iron alloys (Ge 1541 and Armco 18S/R), several commercial oxidation-resistant coatings on AlSl 651 (19-9 DL), and possibly uncoated AISI 310. The best commercial coatings were Cr-Al, Ni-Cr, and a glass ceramic.

ETR, TRA642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF ...

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

ETR, TRA-642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF FLOOR AND FACES NORTH. OUTER WALL OF STORAGE CANAL IS AT RIGHT. SHIELDING IS THICKER AT LOWER LEVEL, WHERE SPENT FUEL ELEMENTS WILL COOL AFTER REMOVAL FROM REACTOR. INL NEGATIVE NO. 56-1401. Jack L. Anderson, Photographer, 5/1/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR CRITICAL FACILITY, TRA654. SCIENTISTS STAND AT EDGE OF TANK ...

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

ETR CRITICAL FACILITY, TRA-654. SCIENTISTS STAND AT EDGE OF TANK AND LIFT REMOVABLE BRIDGE ABOVE THE REACTOR. CONTROL RODS AND FUEL RODS ARE BELOW ENOUGH WATER TO SHIELD WORKERS ABOVE. NOTE CRANE RAILS ALONG WALLS, PUMICE BLOCK WALLS. INL NEGATIVE NO. 57-3690. R.G. Larsen, Photographer, 7/29/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR CONTROL BUILDING, TRA647, INTERIOR. CONTROL ROOM, CONTEXTUAL VIEW. INSTRUMENT ...

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

ETR CONTROL BUILDING, TRA-647, INTERIOR. CONTROL ROOM, CONTEXTUAL VIEW. INSTRUMENT PANELS AT REAR OF OPERATOR'S CONSOLE GAVE OPERATOR STATUS OF REACTOR PERFORMANCE, COOLANT-WATER CHARACTERISTICS AND OTHER INDICATORS. WINDOWS AT RIGHT LOOKED INTO ETR BUILDING FIRST FLOOR. CAMERA FACING EAST. INL NEGATIVE NO. HD42-6. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Electric-stepping-motor tests for a control-drum actuator of a nuclear reactor

NASA Technical Reports Server (NTRS)

Kieffer, A. W.

1972-01-01

Experimental tests were conducted on two stepping motors for application as reactor control-drum actuators. Various control-drum loads with frictional resistances ranging from approximately zero to 40 N-m and inertias ranging from zero to 0.424 kg-sq m were tested.

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

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

Licht, J.; Bergeron, A.; Dionne, B.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimentalmore » device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.« less

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

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

The present plant protection system (PPS) has been defined for use in the TREAT-upgrade (TU) reactor for controlled transient operation of reactor-fuel behavior testing under simulated reactor-accident conditions. A PPS with energy-dependent trip set points lowered worst-case clad temperatures by as much as 180 K, relative to the use of conventional fixed-level trip set points. The multilayered multilevel protection strategy represents the state-of-the-art in terrestrial transient reactor protection systems, and should be applicable to multi-MW space reactors.

Series-Bosch Technology for Oxygen Recovery During Lunar or Martian Surface Missions

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, J. Matthew; Rabenberg, Ellen; Stanley, Christine M.; Edmunson, Jennifer; Alleman, James E.; Chen, Kevin; Dumez, Sam

2014-01-01

Long-duration surface missions to the Moon or Mars will require life support systems that maximize resource recovery to minimize resupply from Earth. To address this need, NASA previously proposed a Series-Bosch (S-Bosch) oxygen recovery system, based on the Bosch process, which can theoretically recover 100% of the oxygen from metabolic carbon dioxide. Bosch processes have the added benefits of the potential to recover oxygen from atmospheric carbon dioxide and the use of regolith materials as catalysts, thereby eliminating the need for catalyst resupply from Earth. In 2012, NASA completed an initial design for an S-Bosch development test stand that incorporates two catalytic reactors in series including a Reverse Water-Gas Shift (RWGS) Reactor and a Carbon Formation Reactor (CFR). In 2013, fabrication of system components, with the exception of a CFR, and assembly of the test stand was initiated. Stand-alone testing of the RWGS reactor was completed to compare performance with design models. Continued testing of Lunar and Martian regolith simulants provided sufficient data to design a CFR intended to utilize these materials as catalysts. Finally, a study was conducted to explore the possibility of producing bricks from spent regolith catalysts. The results of initial demonstration testing of the RWGS reactor, results of continued catalyst performance testing of regolith simulants, and results of brick material properties testing are reported. Additionally, design considerations for a regolith-based CFR are discussed.

Series-Bosch Technology for Oxygen Recovery During Lunar or Martian Surface Missions

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, James M.; Stanley, Christine; Edmunson, Jennifer; Dumez, Samuel; Chen, Kevin; Alleman, James E.

2014-01-01

Long-duration surface missions to the Moon or Mars will require life support systems that maximize resource recovery to minimize resupply from Earth. To address this need, NASA previously proposed a Series-Bosch (S-Bosch) oxygen recovery system, based on the Bosch process, which can theoretically recover 100% of the oxygen from metabolic carbon dioxide. Bosch processes have the added benefits of the potential to recover oxygen from atmospheric carbon dioxide and the use of regolith materials as catalysts, thereby eliminating the need for catalyst resupply from Earth. In 2012, NASA completed an initial design for an S-Bosch development test stand that incorporates two catalytic reactors in series including a Reverse Water-Gas Shift (RWGS) Reactor and a Carbon Formation Reactor (CFR). In 2013, fabrication of system components, with the exception of a CFR, and assembly of the test stand was initiated. Stand-alone testing of the RWGS reactor was completed to compare performance with design models. Continued testing of Lunar and Martian regolith simulants provided sufficient data to design a CFR intended to utilize these materials as catalysts. Finally, a study was conducted to explore the possibility of producing bricks from spend regolith catalysts. The results of initial demonstration testing of the RWGS reactor, results of continued catalyst performance testing of regolith simulants, and results of brick material properties testing are reported. Additionally, design considerations for a regolith-based CFR are discussed.

ETR ELECTRICAL BUILDING, TRA648, INTERIOR. SWITCHGEAR. INL NEGATIVE NO. 563794. ...

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

ETR ELECTRICAL BUILDING, TRA-648, INTERIOR. SWITCHGEAR. INL NEGATIVE NO. 56-3794. Jack L. Anderson, Photographer, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR ELECTRICAL BUILDING, TRA648. BATTERY ROOM. INL NEGATIVE NO. 563785. ...

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

ETR ELECTRICAL BUILDING, TRA-648. BATTERY ROOM. INL NEGATIVE NO. 56-3785. Jack L. Anderson, Photographer, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. INSIDE A FLASH EVAPORATOR. INL NEGATIVE ...

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

PROCESS WATER BUILDING, TRA-605. INSIDE A FLASH EVAPORATOR. INL NEGATIVE NO. 3323. Unknown Photographer, 9/12/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Safety considerations in testing a fuel-rich aeropropulsion gas generator

NASA Technical Reports Server (NTRS)

Rollbuhler, R. James; Hulligan, David D.

1991-01-01

A catalyst containing reactor is being tested using a fuel-rich mixture of Jet A fuel and hot input air. The reactor product is a gaseous fuel that can be utilized in aeropropulsion gas turbine engines. Because the catalyst material is susceptible to damage from high temperature conditions, fuel-rich operating conditions are attained by introducing the fuel first into an inert gas stream in the reactor and then displacing the inert gas with reaction air. Once a desired fuel-to-air ratio is attained, only limited time is allowed for a catalyst induced reaction to occur; otherwise the inert gas is substituted for the air and the fuel flow is terminated. Because there presently is not a gas turbine combustor in which to burn the reactor product gas, the gas is combusted at the outlet of the test facility flare stack. This technique in operations has worked successfully in over 200 tests.

N.S. SAVANNAH, DRAFT OF FINAL SAFEGUARDS REPORT TEST, START-UP AND TRIALS, NEW YORK SHIPBUILDING CORPORATION

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

None

1960-04-01

The N. S. Savannah program for testing, start-up, and initial operation of all reactor and propulsion components and systems is discussed. Definitions of test phases are given and various stages of the test program are outlined. A list of tests for the various reactor, propulsion, and other system components is included. (C.J.G.)

Micro-reactors for characterization of nanostructure-based sensors.

PubMed

Savu, R; Silveira, J V; Flacker, A; Vaz, A R; Joanni, E; Pinto, A C; Gobbi, A L; Santos, T E A; Rotondaro, A L P; Moshkalev, S A

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO(2) substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10(-5) and 10(-1) mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

Micro-reactors for characterization of nanostructure-based sensors

NASA Astrophysics Data System (ADS)

Savu, R.; Silveira, J. V.; Flacker, A.; Vaz, A. R.; Joanni, E.; Pinto, A. C.; Gobbi, A. L.; Santos, T. E. A.; Rotondaro, A. L. P.; Moshkalev, S. A.

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO2 substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10-5 and 10-1 mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

Chemical Reactivity Testing for the National Spent Nuclear Fuel Program. Quality Assurance Project Plan

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

Newsom, H.C.

This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of Work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, QA-101PD, revision 1, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will be noted.

The development and testing of a brief ('gist-based') supplementary colorectal cancer screening information leaflet.

PubMed

Smith, Samuel G; Wolf, Michael S; Obichere, Austin; Raine, Rosalind; Wardle, Jane; von Wagner, Christian

2013-12-01

To design and user-test a 'gist-based' colorectal cancer screening information leaflet, which promotes comprehension of the screening offer. Twenty-eight individuals approaching screening age were recruited from organisations in deprived areas of England. Using a between-subjects design, we tested iterations of a newly-designed gist-based information leaflet. Participants read the leaflet and answered 8 'true' or 'false' comprehension statements. For the leaflet to be considered fit-for-purpose, all statements had to be answered correctly by at least 80% of participants in each round. Alterations were made if this threshold was not met and additional rounds of testing were undertaken. At round 1, answers to 2/8 statements did not meet the threshold. After changes, answers in round 2 did not reach the threshold for 1/8 statements. In round 3, all answers were adequate and the leaflet was deemed fit-for-purpose. Qualitative data offered solutions such as language and layout changes which led to improved comprehension of the leaflet. User-testing substantially improved the design and subsequent comprehensibility of a theory-driven gist-based colorectal cancer screening information leaflet. This leaflet will be evaluated as part of a large national randomised controlled trial designed to reduce socioeconomic inequalities in colorectal cancer screening participation. Copyright © 2013 The Authors. Published by Elsevier Ireland Ltd.. All rights reserved.

What do men understand about lifetime risk following genetic testing? The effect of context and numeracy.

PubMed

Rolison, Jonathan J; Hanoch, Yaniv; Miron-Shatz, Talya

2012-07-01

Genetic testing for gene mutations associated with specific cancers provides an opportunity for early detection, surveillance, and intervention (Smith, Cokkinides, & Brawley, 2008). Lifetime risk estimates provided by genetic testing refer to the risk of developing a specific disease within one's lifetime, and evidence suggests that this is important for the medical choices people make, as well as their future family and financial plans. The present studies tested whether adult men understand the lifetime risks of prostate cancer informed by genetic testing. In 2 experiments, adult men were asked to interpret the lifetime risk information provided in statements about risks of prostate cancer. Statement format was manipulated such that the most appropriate interpretation of risk statements referred to an absolute risk of cancer in experiment 1 and a relative risk in experiment 2. Experiment 1 revealed that few men correctly interpreted the lifetime risks of cancer when these refer to an absolute risk of cancer, and numeracy levels positively predicted correct responding. The proportion of correct responses was greatly improved in experiment 2 when the most appropriate interpretation of risk statements referred instead to a relative rather than an absolute risk, and numeracy levels were less involved. Understanding of lifetime risk information is often poor because individuals incorrectly believe that these refer to relative rather than absolute risks of cancer.

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

NASA Astrophysics Data System (ADS)

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

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

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

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

Woods, Brian; Gutowska, Izabela; Chiger, Howard

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

Plum Brook Reactor Facility Control Room during Facility Startup

NASA Image and Video Library

1961-02-21

Operators test the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility systems in the months leading up to its actual operation. The “Reactor On” signs are illuminated but the reactor core was not yet ready for chain reactions. Just a couple weeks after this photograph, Plum Brook Station held a media open house to unveil the 60-megawatt test reactor near Sandusky, Ohio. More than 60 members of the print media and radio and television news services met at the site to talk with community leaders and representatives from NASA and Atomic Energy Commission. The Plum Brook reactor went critical for the first time on the evening of June 14, 1961. It was not until April 1963 that the reactor reached its full potential of 60 megawatts. The reactor control room, located on the second floor of the facility, was run by licensed operators. The operators manually operated the shim rods which adjusted the chain reaction in the reactor core. The regulating rods could partially or completely shut down the reactor. The control room also housed remote area monitoring panels and other monitoring equipment that allowed operators to monitor radiation sensors located throughout the facility and to scram the reactor instantly if necessary. The color of the indicator lights corresponded with the elevation of the detectors in the various buildings. The reactor could also shut itself down automatically if the monitors detected any sudden irregularities.

Research Program of a Super Fast Reactor

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

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

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

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

DOE PAGES

Melin, Alexander M.; Kisner, Roger A.

2018-04-03

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

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

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

Melin, Alexander M.; Kisner, Roger A.

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

PBF Reactor Building (PER620). Fuel rod test assembly is on ...

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

PBF Reactor Building (PER-620). Fuel rod test assembly is on display at PBF. Date: 1982. INEEL negative no. 82-4893 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION ...

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

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION EXPERIMENT NO. 1. INL NEGATIVE NO. 6510. Unknown Photographer, 9/29/1959 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PRECAST CONCRETE WALL PANELS ARE LIFTED INTO PLACE ON MTR ...

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

PRECAST CONCRETE WALL PANELS ARE LIFTED INTO PLACE ON MTR STEEL FRAME STRUCTURE. INL NEGATIVE NO. 1330. Unknown Photographer, 1/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

FAST CHOPPER BUILDING, TRA665, INTERIOR. UPPER LEVEL. CONCRETE WALLS. INL ...

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

FAST CHOPPER BUILDING, TRA-665, INTERIOR. UPPER LEVEL. CONCRETE WALLS. INL NEGATIVE NO. HD42-2. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

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

Diamond, D. J.; Varuttamaseni, A.

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

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

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

D. J. Diamond

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

10 CFR 50.58 - Hearings and report of the Advisory Committee on Reactor Safeguards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Hearings and report of the Advisory Committee on Reactor... Hearings and report of the Advisory Committee on Reactor Safeguards. (a) Each application for a....22, or for a testing facility, shall be referred to the Advisory Committee on Reactor Safeguards for...

10 CFR 50.58 - Hearings and report of the Advisory Committee on Reactor Safeguards.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Hearings and report of the Advisory Committee on Reactor... Hearings and report of the Advisory Committee on Reactor Safeguards. (a) Each application for a....22, or for a testing facility, shall be referred to the Advisory Committee on Reactor Safeguards for...

Cost-Effective Systems for Atomic Layer Deposition

ERIC Educational Resources Information Center

Lubitz, Michael; Medina, Phillip A., IV; Antic, Aleks; Rosin, Joseph T.; Fahlman, Bradley D.

2014-01-01

Herein, we describe the design and testing of two different home-built atomic layer deposition (ALD) systems for the growth of thin films with sub-monolayer control over film thickness. The first reactor is a horizontally aligned hot-walled reactor with a vacuum purging system. The second reactor is a vertically aligned cold-walled reactor with a…

Simulation of Watts Bar Unit 1 Initial Startup Tests with Continuous Energy Monte Carlo Methods

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

Godfrey, Andrew T; Gehin, Jess C; Bekar, Kursat B

2014-01-01

The Consortium for Advanced Simulation of Light Water Reactors* is developing a collection of methods and software products known as VERA, the Virtual Environment for Reactor Applications. One component of the testing and validation plan for VERA is comparison of neutronics results to a set of continuous energy Monte Carlo solutions for a range of pressurized water reactor geometries using the SCALE component KENO-VI developed by Oak Ridge National Laboratory. Recent improvements in data, methods, and parallelism have enabled KENO, previously utilized predominately as a criticality safety code, to demonstrate excellent capability and performance for reactor physics applications. The highlymore » detailed and rigorous KENO solutions provide a reliable nu-meric reference for VERAneutronics and also demonstrate the most accurate predictions achievable by modeling and simulations tools for comparison to operating plant data. This paper demonstrates the performance of KENO-VI for the Watts Bar Unit 1 Cycle 1 zero power physics tests, including reactor criticality, control rod worths, and isothermal temperature coefficients.« less

Risk factors for bovine tuberculosis in low incidence regions related to the movements of cattle.

PubMed

Gates, M Carolyn; Volkova, Victoriya V; Woolhouse, Mark E J

2013-11-09

Bovine tuberculosis (bTB) remains difficult to eradicate from low incidence regions partly due to the imperfect sensitivity and specificity of routine intradermal tuberculin testing. Herds with unconfirmed reactors that are incorrectly classified as bTB-negative may be at risk of spreading disease, while those that are incorrectly classified as bTB-positive may be subject to costly disease eradication measures. This analysis used data from Scotland in the period leading to Officially Tuberculosis Free recognition (1) to investigate the risks associated with the movements of cattle from herds with different bTB risk classifications and (2) to identify herd demographic characteristics that may aid in the interpretation of tuberculin testing results. From 2002 to 2009, for every herd with confirmed bTB positive cattle identified through routine herd testing, there was an average of 2.8 herds with at least one unconfirmed positive reactor and 18.9 herds with unconfirmed inconclusive reactors. Approximately 75% of confirmed bTB positive herds were detected through cattle with no known movements outside Scotland. At the animal level, cattle that were purchased from Scottish herds with unconfirmed positive reactors and a recent history importing cattle from endemic bTB regions were significantly more likely to react positively on routine intradermal tuberculin tests, while cattle purchased from Scottish herds with unconfirmed inconclusive reactors were significantly more likely to react inconclusively. Case-case comparisons revealed few demographic differences between herds with confirmed positive, unconfirmed positive, and unconfirmed inconclusive reactors, which highlights the difficulty in determining the true disease status of herds with unconfirmed tuberculin reactors. Overall, the risk of identifying reactors through routine surveillance decreased significantly over time, which may be partly attributable to changes in movement testing regulations and the volume of cattle imported from endemic regions. Although the most likely source of bTB infections in Scotland was cattle previously imported from endemic regions, we found indirect evidence of transmission within Scottish cattle farms and cannot rule out the possibility of low level transmission between farms. Further investigation is needed to determine whether targeting herds with unconfirmed reactors and a history of importing cattle from high risk regions would benefit control efforts.

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

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

McCullough, C.R.

1958-10-31

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

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

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

Honma, George

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will bemore » used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-18

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

A 640 foot per second impact test of a two foot diameter model nuclear reactor containment system without fracture

NASA Technical Reports Server (NTRS)

Puthoff, R. L.

1971-01-01

An impact test was conducted on an 1142 pound 2 foot diameter sphere model. The purpose of this test was to determine the feasibility of containing the fission products of a mobile reactor in an impact. The model simulated the reactor core, energy absorbing gamma shielding, neutron shielding and the containment vessel. It was impacted against an 18,000 pound reinforced concrete block. The model was significantly deformed and the concrete block demolished. No leaks were detected nor cracks observed in the model after impact.

High velocity continuous-flow reactor for the production of solar grade silicon

NASA Technical Reports Server (NTRS)

Woerner, L.

1977-01-01

The feasibility of a high volume, high velocity continuous reduction reactor as an economical means of producing solar grade silicon was tested. Bromosilanes and hydrogen were used as the feedstocks for the reactor along with preheated silicon particles which function both as nucleation and deposition sites. A complete reactor system was designed and fabricated. Initial preheating studies have shown the stability of tetrabromosilane to being heated as well as the ability to preheat hydrogen to the desired temperature range. Several test runs were made and some silicon was obtained from runs carried out at temperatures in excess of 1180 K.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

NASA Astrophysics Data System (ADS)

Koshelev, A. S.; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

2016-12-01

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

[Radiation ecological environment in the Republic of Kazakhstan in the vicinity of the reactors and on the territory of the Semipalatinsk Test Site].

PubMed

Kim, D S

2012-01-01

The results of research into the environmental conditions in the regions of location of the pressurized water reactor WWR-K, fast neutron breeder BN-350 and on the territory of the Semipalatinsk Test Site are represented. The effects of the exposure to aerosol emissions from WWR-K and BN-350 reactors on the environment are summarized. We present some arguments in favor of the safe operation of fission reactors in compliance with the rules and norms of nuclear and radiation protection and the efficient disposal of radioactive waste on the territory of the Republic.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

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

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Kovshov, K. N.; Ovchinnikov, M. A.

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

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

Jennifer Lyons; Wade R. Marcum; Mark D. DeHart

2014-01-01

The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by themore » Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.« less

Corrosion resistance of 0Kh18N10T steel in gadolinium nitrate solutions in the liquid regulation of the reactivity of nuclear reactors

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

Ganzha, V.D.; Konoplev, K.A.; Mashchetov, V.P.

1986-03-01

This study was carried out in connection with the preparation of the design for the PIK research reactor. The corrosion resistance of 0Kh18N10T steel in gadolinium nitrate solutions was tested in laboratory, ampule, and loop corrosion tests. At all stages of the tests, the authors investigated the effect produced on the corrosion processes by factors related to the technology of preparation of the equipment (mechanical working of the surfaces, welding, sensitizing, annealing, stressed state of the material, cracks, etc.). Ampule tests were conducted in order to determine the effect produced by reactor radiation and shutdown regimes on the corrosion resistancemore » of the steel. Special ampules made of 0Kh18N10T steel were filled with gadolinium nitrate solutions of various concentrations, sealed, and irradiated for a long period in the core of the VVR-M reactor at a temperature of 20-50 degrees C. The results of the tests are shown. The investigations showed that the corrosion of 0Kh18N10T steel in solutions of gadolinium nitrate is uniform, regardless of the state of the surface, the concentration of gadolinium nitrate, the duration of the tests, the action of the reactor radiation under static and dynamic conditions, and the presence of mechanical stresses.« less

United States and Russian Cooperation on Issues of Nuclear Nonproliferation

DTIC Science & Technology

2005-06-01

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

IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ...

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

IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ion chamber beneath reactor on dolly. Photographer: Page Comiskey. Date: August 11, 1965. INEEL negative no. 65-4039 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

DTIC Science & Technology

1987-06-01

computer modeling remains at best semiempirical (C-i), this large variation in scaling factor makes extrapolation of data impossible. The DIDO Water...in a full scale PWR are not practical. The reactor plant is not controlled to tolerances necessary for research, and utilities are reluctant to vary...MIT Reactor Safeguards Committee, in revision 1 to the PCCL Safety Evaluation Report (SER), for final approval to begin in-pile testing and

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

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

Rohatgi, Upendra S.

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary ofmore » appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/« less

STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

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

Dickerman, C. E.; Sowa, E. S.; Okrent, D.

1961-08-01

Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

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

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

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

2007-07-01

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

The Society for Obstetric Anesthesia and Perinatology consensus statement on the management of cardiac arrest in pregnancy.

PubMed

Lipman, Steven; Cohen, Sheila; Einav, Sharon; Jeejeebhoy, Farida; Mhyre, Jill M; Morrison, Laurie J; Katz, Vern; Tsen, Lawrence C; Daniels, Kay; Halamek, Louis P; Suresh, Maya S; Arafeh, Julie; Gauthier, Dodi; Carvalho, Jose C A; Druzin, Maurice; Carvalho, Brendan

2014-05-01

This consensus statement was commissioned in 2012 by the Board of Directors of the Society for Obstetric Anesthesia and Perinatology to improve maternal resuscitation by providing health care providers critical information (including point-of-care checklists) and operational strategies relevant to maternal cardiac arrest. The recommendations in this statement were designed to address the challenges of an actual event by emphasizing health care provider education, behavioral/communication strategies, latent systems errors, and periodic testing of performance. This statement also expands on, interprets, and discusses controversial aspects of material covered in the American Heart Association 2010 guidelines.

Tower Shielding Reactor II design and operation report: Vol. 2. Safety Analysis

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

Holland, L. B.; Kolb, J. O.

1970-01-01

Information on the Tower Shielding Reactor II is contained in the TSR-II Design and Operation Report and in the Tower Shielding Facility Manual. The TSR-II Design and Operating Report consists of three volumes. Volume 1 is Descriptions of the Tower Shielding Reactor II and Facility; Volume 2 is Safety analysis of the Tower Shielding Reactor II; and Volume 3 is the Assembly and Testing of the Tower Shielding Reactor II Control Mechanism Housing.

ADVANCED HEAT TRANSFER TEST FACILITY, TRA666A. ELEVATIONS. ROOF FRAMING PLAN. ...

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

ADVANCED HEAT TRANSFER TEST FACILITY, TRA-666A. ELEVATIONS. ROOF FRAMING PLAN. CONCRETE BLOCK SIDING. SLOPED ROOF. ROLL-UP DOOR. AIR INTAKE ENCLOSURE ON NORTH SIDE. F.C. TORKELSON 842-MTR-666-A5, 8/1966. INL INDEX NO. 531-0666-00-851-152258, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Fuel processing in integrated micro-structured heat-exchanger reactors

NASA Astrophysics Data System (ADS)

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

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

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

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

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

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

'If you are empathetic you care about both animals and people. I am a nurse and I don't like to see suffering anywhere': Findings from 103 healthcare professionals on attitudes to animal experimentation.

PubMed

Dignon, Andrée

2016-11-01

This report presents qualitative and quantitative data from 103 UK healthcare professionals describing attitudes to the current system of animal testing (to produce medicines and health interventions). To gather qualitative testimony, these healthcare professionals were organised into six separate focus groups (of 18, 17, 17, 15, 17 and 19 participants) where they were asked 'what is your opinion about the current system of animal testing?' The study focussed on attitudes to the current system rather than attitudes to animal testing in general. The healthcare professionals also completed a quantitative attitude scale questionnaire consisting of 20 statements (all favourable) towards the system of animal testing as currently practised. Statements such as 'Testing agencies abide by legislation to safeguard animal welfare' were displayed and the healthcare professionals were invited to agree or disagree with these statements. The results from both the quantitative and qualitative data suggest that healthcare professionals were opposed to the current system of animal experimentation.

MTR CONTROL ROOM WITH CONTROL CONSOLE AND STATUS READOUTS ALONG ...

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

MTR CONTROL ROOM WITH CONTROL CONSOLE AND STATUS READOUTS ALONG WALL. WORKERS MAKE ELECTRICAL AND OTHER CONNECTIONS. INL NEGATIVE NO. 4289. Unknown Photographer, 2/26/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

10 CFR 100.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... approval requirements for proposed sites for stationary power and testing reactors subject to part 50 or part 52 of this chapter. (b) There exists a substantial base of knowledge regarding power reactor... approach incorporates the appropriate standards and criteria for approval of stationary power and testing...

SOUTH WING, TRA661. SOUTH SIDE. CAMERA FACING NORTH. MTR HIGH ...

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

SOUTH WING, TRA-661. SOUTH SIDE. CAMERA FACING NORTH. MTR HIGH BAY BEYOND. INL NEGATIVE NO. HD46-45-3. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR HEAT EXCHANGER BUILDING, TRA644. WORKERS ARE INSTALLING HEAT EXCHANGER ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. WORKERS ARE INSTALLING HEAT EXCHANGER PIPING. INL NEGATIVE NO. 56-3122. Jack L. Anderson, Photographer, 9/21/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR AND ETR COMPLEXES. CAMERA FACING EASTERLY TOWARD CHEMICAL PROCESSING ...

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

MTR AND ETR COMPLEXES. CAMERA FACING EASTERLY TOWARD CHEMICAL PROCESSING PLANT. MTR AND ITS ATTACHMENTS IN FOREGROUND. ETR BEYOND TO RIGHT. INL NEGATIVE NO. 56-4100. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR COMPRESSOR BUILDING, TRA643. COMPRESSORS AND OTHER EQUIPMENT INSTALLED. METAL ...

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

ETR COMPRESSOR BUILDING, TRA-643. COMPRESSORS AND OTHER EQUIPMENT INSTALLED. METAL ROOF AND CONCRETE BLOCK WALLS. INL NEGATIVE NO. 61-4536. Unknown Photographer, ca. 1961. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

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

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

1981-04-01

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

29 CFR 1910.1096 - Ionizing radiation.

Code of Federal Regulations, 2010 CFR

2010-07-01

... test tubes, used transiently in laboratory procedures, when the user is present. (iv) Where containers... shall contain the following statement: “You should preserve this report for future reference.” (n... statement: “You should preserve this report for future reference.” (2) [Reserved] (p) Nuclear Regulatory...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2011 CFR

2011-01-01

.... Containment inspection. B. Repordkeeping of test results. I. Introduction One of the conditions of all... following: A. Type A test—1. Pretest requirements. (a) Containment inspection in accordance with V. A. shall.... During the period between the completion of one Type A test and the initiation of the containment...

PBF Reactor Building (PER620). Detail of fuel test assembly in ...

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

PBF Reactor Building (PER-620). Detail of fuel test assembly in preparation for test. When complete, it will fit into in-pile tube. The maximum outside diameter of which must be about 8.25 inches. Date: 1982. INEEL negative no. 82-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Research and Test Reactors.'' This guide describes a method that the staff of the NRC considers acceptable... assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in research...

Spirometry in primary care

PubMed Central

Coates, Allan L; Graham, Brian L; McFadden, Robin G; McParland, Colm; Moosa, Dilshad; Provencher, Steeve; Road, Jeremy

2013-01-01

Canadian Thoracic Society (CTS) clinical guidelines for asthma and chronic obstructive pulmonary disease (COPD) specify that spirometry should be used to diagnose these diseases. Given the burden of asthma and COPD, most people with these diseases will be diagnosed in the primary care setting. The present CTS position statement was developed to provide guidance on key factors affecting the quality of spirometry testing in the primary care setting. The present statement may also be used to inform and guide the accreditation process for spirometry in each province. Although many of the principles discussed are equally applicable to pulmonary function laboratories and interpretation of tests by respirologists, they are held to a higher standard and are outside the scope of the present statement. PMID:23457669

Imaging Fukushima Daiichi reactors with muons

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

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.

2013-05-15

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi tomore » make this determination in the near future.« less

Imaging Fukushima Daiichi reactors with muons

NASA Astrophysics Data System (ADS)

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Lukić, Zarija; Masuda, Koji; Milner, Edward C.; Morris, Christopher L.; Perry, John O.

2013-05-01

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

Space Nuclear Reactor Engineering

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

Poston, David Irvin

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

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

The portrayal of Down syndrome in prenatal screening information pamphlets.

PubMed

Lawson, Karen L; Carlson, Kara; Shynkaruk, Jody M

2012-08-01

To examine the information about Down syndrome (DS) provided to pregnant women in Canada through a content analysis of prenatal screening information pamphlets. Prenatal screening information pamphlets were requested from Canadian prenatal testing centres. In total, 17 pamphlets were received (response rate = 65%). Statements presenting information descriptive of DS were identified from the pamphlets, and a content analysis was carried out. Specifically, each statement was analyzed with respect to both the content and the valence of the information presented on the basis of predetermined decision rules. To enhance reliability, four independent raters reviewed each statement, and any differences in coding were resolved through discussion. In total, 158 statements descriptive of DS were extracted from the pamphlets. The categorical analysis revealed that 91% of the extracted statements emphasized medical or clinical information about DS, whereas only 9% of the statements relayed information pertaining to psychosocial issues. The valence analysis revealed that nearly one half of the statements portrayed a negative message pertaining to DS, while only 2.4% of the statements conveyed a positive image of DS. The pamphlets provided to pregnant women do not appear to present a comprehensive, balanced portrayal of DS, which may serve to limit informed decision-making.

Assessment of Sensor Technologies for Advanced Reactors

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

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

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

Current and prospective safety issues at the HFBR

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

Tichler, P.R.

The Brookhaven high-flux beam reactor (HFBR) was designed primarily to produce external neutron beams for experimental research. It is cooled, moderated, and reflected by heavy water and uses materials test reactor and engineering test reactor type of fuel elements containing enriched uranium. The reactor power when operation began in 1965 was 40 MW, was raised to 60 MW in 1982 after a number of plant modifications, and operated at that level until 1989. Since that time, safety questions have been raised that resulted in extended shutdowns and a reduction in operating power to 30 MW. This paper discusses the principalmore » safety issues and plans for their resolution and return to 60-MW operation. In addition, radiation embrittlement of the reactor vessel and thermal shield and its effect on the life of the facility are briefly discussed.« less

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA603. SUMMARY OF ...

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

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA-603. SUMMARY OF COOLANT FLOW FROM WORKING RESERVOIR TO INTERIOR OF REACTOR'S THERMAL SHIELD. NAMES TANK SECTIONS. PIPE AND DRAIN-LINE SIZES. SHOWS DIRECTION OF AIR FLOW THROUGH PEBBLE AND GRAPHITE BLOCK ZONE. NEUTRON CURTAIN AND THERMAL COLUMN DOOR. BLAW-KNOX 3150-92-7, 3/1950. INL INDEX NO. 531-0603-51-098-100036, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETRMTR MECHANICAL SERVICES BUILDING, TRA653. CAMERA FACING NORTHWEST AS BUILDING ...

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

ETR-MTR MECHANICAL SERVICES BUILDING, TRA-653. CAMERA FACING NORTHWEST AS BUILDING WAS NEARLY COMPLETE. INL NEGATIVE NO. 57-3653. K. Mansfield, Photographer, 7/22/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

DESIGN CRITERIA FOR HIGH TEMPERATURE LATTICE TEST REACTOR PROJECT CAH-100

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

Ballard, D.L.; Brown, W.W.; Harrison, C.W.

Design and construction specifications to be followed in the development of the reactor, its associated systems and experimental facilities, and the housing and required services for the facility are presented. The testing procedures to be used are outlined. (D.C.W.)

FAST CHOPPER BUILDING, TRA665. DETAIL OF STEEL DOOR ENTRY TO ...

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

FAST CHOPPER BUILDING, TRA-665. DETAIL OF STEEL DOOR ENTRY TO LOWER LEVEL. CAMERA FACING NORTH. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PBF Reactor Building (PER620). Camera on main floor faces south ...

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

PBF Reactor Building (PER-620). Camera on main floor faces south (open) doorway. In foreground is canal gate, lined with stainless steel and painted with protective coatings. Reactor pit is round with protective coatings. Reactor put is round form discernible beyond. Lifting beams and rigging are in place for a load test before reactor vessel arrives. Photographer: John Capek. Date: January 26, 1970. INEEL negative no. 70-347 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Pilot plant operation of a nonadiabatic methanation reactor. [15 refs. ; Raney nickel catalyst

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

Schehl, R.R.; Pennline, H.W.; Strakey, J.P.

The design and operation of a pilot plant scale hybrid methanation reactor is discussed. The hybrid methanator, utilizing a finned, Raney nickel coated insert, consolidates features of the tube-wall and hot-gas-recycle methanation reactors. Data are presented from four tests lasting from 3/sup 1///sub 2/ weeks to three months. Topics discussed include conversion, product yields, catalyst properties, and reactor temperature profiles. A one-dimensional mathematical model capable of explaining reactor performance trends is employed.

Hodoscope Cineradiography Of Nuclear Fuel Destruction Experiments

NASA Astrophysics Data System (ADS)

De Volpi, A.

1983-08-01

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

Shear compression testing of glass-fibre steel specimens after 4K reactor irradiation: Present status and facility upgrade

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

Gerstenberg, H.; Kraehling, E.; Katheder, H.

1997-06-01

The shear strengths of various fibre reinforced resins being promising candidate insulators for superconducting coils to be used tinder a strong radiation load, e.g. in future fusion reactors were investigated prior and subsequent to reactor in-core irradiation at liquid helium temperature. A large number of sandwich-like (steel-bonded insulation-steel) specimens representing a widespread variety of materials and preparation techniques was exposed to irradiation doses of up to 5 x 10{sup 7} Gy in form of fast neutrons and {gamma}-radiation. In a systematic study several experimental parameters including irradiation dose, postirradiation storage temperature and measuring temperature were varied before the determination ofmore » the ultimate shear strength. The results obtained from the different tested materials are compared. In addition an upgrade of the in-situ test rig installed at the Munich research reactor is presented, which allows combined shear/compression loading of low temperature irradiated specimens and provides a doubling of the testing rate.« less

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

Hyperthermal Environments Simulator for Nuclear Rocket Engine Development

NASA Technical Reports Server (NTRS)

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

2011-01-01

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

Closed Brayton cycle power conversion systems for nuclear reactors :

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

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

2006-04-01

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

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

DTIC Science & Technology

2008-10-17

safeguards-irrelevant.” The following facilities and activities were not on the separation list: ! 8 indigenous Indian power reactors ! Fast Breeder ...test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction ! Enrichment facilities ! Spent fuel reprocessing facilities (except...potential use in a bomb. In addition, safeguards on enrichment, reprocessing plants, and breeder reactors would support the 2002 U.S. National Strategy to

Quantifying Acoustic Impacts on Marine Mammals and Sea Turtles: Methods and Analytical Approach for Phase III Training and Testing

DTIC Science & Technology

2017-06-16

Acoustic Impacts on Marine Mammals and Sea Turtles: Methods and Analytical Approach for Phase III Training and Testing Sarah A. Blackstock Joseph O...December 2017 4. TITLE AND SUBTITLE Quantifying Acoustic Impacts on Marine Mammals and Sea Turtles: Methods and Analytical Approach for Phase III...Navy’s Phase III Study Areas as described in each Environmental Impact Statement/ Overseas Environmental Impact Statement and describes the methods

Nuclear fuel management optimization using genetic algorithms

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

DeChaine, M.D.; Feltus, M.A.

1995-07-01

The code independent genetic algorithm reactor optimization (CIGARO) system has been developed to optimize nuclear reactor loading patterns. It uses genetic algorithms (GAs) and a code-independent interface, so any reactor physics code (e.g., CASMO-3/SIMULATE-3) can be used to evaluate the loading patterns. The system is compared to other GA-based loading pattern optimizers. Tests were carried out to maximize the beginning of cycle k{sub eff} for a pressurized water reactor core loading with a penalty function to limit power peaking. The CIGARO system performed well, increasing the k{sub eff} after lowering the peak power. Tests of a prototype parallel evaluation methodmore » showed the potential for a significant speedup.« less

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

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

Konzek, G.J.

1983-07-01

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.

MTR MAIN FLOOR. NEUTRON TUNNEL (SPANNED BY STILELIKE STEPS) PROJECTS ...

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

MTR MAIN FLOOR. NEUTRON TUNNEL (SPANNED BY STILE-LIKE STEPS) PROJECTS FROM THE SOUTHEAST CORNER OF THE MTR TOWARD SOUTHEAST CORNER OF BUILDING, WHERE SHIELDING BLOCKS BEGIN TO SURROUND THE TUNNEL AS IT NEARS DETECTING INSTRUMENTS NEAR THE BUILDING WALL. GEAR RELATED TO CRYSTAL NEUTRON SPECTROMETER IS IN FOREGROUND SURROUNDED BY SHIELDING. DATA CONSOLES ARE AT MID-LEVEL OF EAST FACE. OTHER WORK PROCEEDS ON TOP OF AND ELSEWHERE AROUND REACTOR. NOTE TOOLS HANGING AGAINST SOUTHEAST CORNER, USED TO CHANGE FUEL ELEMENTS AND OTHER REACTOR ITEMS DURING REFUELING CYCLES. INL NEGATIVE NO. 10439. Unknown Photographer, 4/20/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

TEST-HOLE CONSTRUCTION FOR A NEUTRONIC REACTOR

DOEpatents

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

1959-02-17

Test-hole construction is described for a reactor which provides safe and ready access to the neutron flux region for specimen materials which are to be irradiated therein. An elongated tubular thimble adapted to be inserted in the access hole through the wall of the reactor is constructed of aluminum and is provided with a plurality of holes parallel to the axis of the thimble for conveying the test specimens into position for irradiation, and a conduit for the circulation of coolant. A laminated shield formed of alternate layers of steel and pressed wood fiber is disposed lengthwise of the thimble near the outer end thereof.

Reduced enrichment for research and test reactors: Proceedings

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

Not Available

1993-07-01

The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

DOE PAGES

Kim, Ki-Hwan; Kim, Jong-Hwan; Oh, Seok-Jin; ...

2016-01-01

The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the developmentmore » of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.« less

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

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

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

2016-04-01

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

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

Lisowski, Darius D.; Kraus, Adam R.; Bucknor, Matthew D.

A 1/2 scale test facility has been constructed at Argonne National Laboratory to study the heat removal performance and natural circulation flow patterns in a Reactor Cavity Cooling System (RCCS). Our test facility, the Natural convection Shutdown heat removal Test Facility (NSTF), supports the broader goal of developing an inherently safe and fully passive ex-vessel decay heat removal for advanced reactor designs. The project, initiated in 2010 to support the Advanced Reactor Technologies (ART), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs, has been conducting experimental operations since early 2014. The following paper provides a summary ofmore » some primary design features of the 26-m tall test facility along with a description of the data acquisition suite that guides our experimental practices. Specifics of the distributed fiber optic temperature measurements will be discussed, which introduces an unparalleled level of data density that has never before been implemented in a large scale natural circulation test facility. Results from our test series will then be presented, which provide insight into the thermal hydraulic behavior at steady-state and transient conditions for varying heat flux levels and exhaust chimney configuration states. (C) 2016 Elsevier B.V. All rights reserved.« less

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

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

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated meltingmore » of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.« less

SAFETY AND SECURITY BUILDING, TRA614. ELEVATIONS. SECTIONS. TWO ROOF LEVELS. ...

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

SAFETY AND SECURITY BUILDING, TRA-614. ELEVATIONS. SECTIONS. TWO ROOF LEVELS. BLAW-KNOX 3150-814-2, 3/1950. INL INDEX NO. 531-0614-00-098-100703, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

SAFETY AND SECURITY BUILDING, TRA614. SIMPLIFIED FLOOR LAYOUT AND WEST ...

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

SAFETY AND SECURITY BUILDING, TRA-614. SIMPLIFIED FLOOR LAYOUT AND WEST ELEVATION. BLAW-KNOX 3150-14-1, 1/1950. INL INDEX NO. 531-0614-00-098-100024, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. FLASH EVAPORATOR, CONDENSER (PROJECT FROM EVAPORATOR), ...

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

PROCESS WATER BUILDING, TRA-605. FLASH EVAPORATOR, CONDENSER (PROJECT FROM EVAPORATOR), AND STEAM EJECTOR (ALONG REAR WALL). INL NEGATIVE NO. 4377. M.H. Bartz, Photographer, 3/5/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

FAN HOUSE INTERIOR. THREE MOTOR DRIVES FOR POSITIVE DISPLACEMENT BLOWERS ...

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

FAN HOUSE INTERIOR. THREE MOTOR DRIVES FOR POSITIVE DISPLACEMENT BLOWERS LINE UP ON NORTH WALL. CONCRETE PEDESTALS. CAMERA FACES NORTHEAST. INL NEGATIVE NO. 4291. Unknown Photographer, 2/26/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR BUILDING AND BALCONY FLOORS. CAMERA FACING EASTERLY. PHOTOGRAPHER DID ...

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

MTR BUILDING AND BALCONY FLOORS. CAMERA FACING EASTERLY. PHOTOGRAPHER DID NOT EXPLAIN DARK CLOUD. MTR WING WILL ATTACH TO GROUND FLOOR. INL NEGATIVE NO. 1567. Unknown Photographer, 2/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE SHOWN IN ID-33-G-101. INL NEGATIVE NO. HD24-3-4. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The development of a universal diagnostic probe system for Tokamak fusion test reactor

NASA Technical Reports Server (NTRS)

Mastronardi, R.; Cabral, R.; Manos, D.

1982-01-01

The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

10 CFR 100.1 - Purpose.

Code of Federal Regulations, 2011 CFR

2011-01-01

... COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.1 Purpose. (a) The purpose of this part is to establish approval requirements for proposed sites for stationary power and testing reactors subject to part 50 or part 52 of this chapter. (b) There exists a substantial base of knowledge regarding power reactor...

9 CFR 309.14 - Brucellosis-reactor goats.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 9 Animals and Animal Products 2 2014-01-01 2014-01-01 false Brucellosis-reactor goats. 309.14... INSPECTION AND CERTIFICATION ANTE-MORTEM INSPECTION § 309.14 Brucellosis-reactor goats. Goats which have reacted to a test for brucellosis shall not be slaughtered in an official establishment. ...

9 CFR 309.14 - Brucellosis-reactor goats.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 9 Animals and Animal Products 2 2013-01-01 2013-01-01 false Brucellosis-reactor goats. 309.14... INSPECTION AND CERTIFICATION ANTE-MORTEM INSPECTION § 309.14 Brucellosis-reactor goats. Goats which have reacted to a test for brucellosis shall not be slaughtered in an official establishment. ...

9 CFR 309.14 - Brucellosis-reactor goats.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 9 Animals and Animal Products 2 2012-01-01 2012-01-01 false Brucellosis-reactor goats. 309.14... INSPECTION AND CERTIFICATION ANTE-MORTEM INSPECTION § 309.14 Brucellosis-reactor goats. Goats which have reacted to a test for brucellosis shall not be slaughtered in an official establishment. ...

Preferred Reporting Items for a Systematic Review and Meta-analysis of Diagnostic Test Accuracy Studies: The PRISMA-DTA Statement.

PubMed

McInnes, Matthew D F; Moher, David; Thombs, Brett D; McGrath, Trevor A; Bossuyt, Patrick M; Clifford, Tammy; Cohen, Jérémie F; Deeks, Jonathan J; Gatsonis, Constantine; Hooft, Lotty; Hunt, Harriet A; Hyde, Christopher J; Korevaar, Daniël A; Leeflang, Mariska M G; Macaskill, Petra; Reitsma, Johannes B; Rodin, Rachel; Rutjes, Anne W S; Salameh, Jean-Paul; Stevens, Adrienne; Takwoingi, Yemisi; Tonelli, Marcello; Weeks, Laura; Whiting, Penny; Willis, Brian H

2018-01-23

Systematic reviews of diagnostic test accuracy synthesize data from primary diagnostic studies that have evaluated the accuracy of 1 or more index tests against a reference standard, provide estimates of test performance, allow comparisons of the accuracy of different tests, and facilitate the identification of sources of variability in test accuracy. To develop the Preferred Reporting Items for Systematic Reviews and Meta-Analyses (PRISMA) diagnostic test accuracy guideline as a stand-alone extension of the PRISMA statement. Modifications to the PRISMA statement reflect the specific requirements for reporting of systematic reviews and meta-analyses of diagnostic test accuracy studies and the abstracts for these reviews. Established standards from the Enhancing the Quality and Transparency of Health Research (EQUATOR) Network were followed for the development of the guideline. The original PRISMA statement was used as a framework on which to modify and add items. A group of 24 multidisciplinary experts used a systematic review of articles on existing reporting guidelines and methods, a 3-round Delphi process, a consensus meeting, pilot testing, and iterative refinement to develop the PRISMA diagnostic test accuracy guideline. The final version of the PRISMA diagnostic test accuracy guideline checklist was approved by the group. The systematic review (produced 64 items) and the Delphi process (provided feedback on 7 proposed items; 1 item was later split into 2 items) identified 71 potentially relevant items for consideration. The Delphi process reduced these to 60 items that were discussed at the consensus meeting. Following the meeting, pilot testing and iterative feedback were used to generate the 27-item PRISMA diagnostic test accuracy checklist. To reflect specific or optimal contemporary systematic review methods for diagnostic test accuracy, 8 of the 27 original PRISMA items were left unchanged, 17 were modified, 2 were added, and 2 were omitted. The 27-item PRISMA diagnostic test accuracy checklist provides specific guidance for reporting of systematic reviews. The PRISMA diagnostic test accuracy guideline can facilitate the transparent reporting of reviews, and may assist in the evaluation of validity and applicability, enhance replicability of reviews, and make the results from systematic reviews of diagnostic test accuracy studies more useful.

FFTF Passive Safety Test Data for Benchmarks for New LMR Designs

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

Wootan, David W.; Casella, Andrew M.

Liquid Metal Reactors (LMRs) continue to be considered as an attractive concept for advanced reactor design. Software packages such as SASSYS are being used to im-prove new LMR designs and operating characteristics. Significant cost and safety im-provements can be realized in advanced liquid metal reactor designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associ-ated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. The FFTF passive safety testing pro-gram was developed to examine howmore » specific design elements influenced dynamic re-activity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results to reactors of current interest. The U.S. Department of En-ergy, Office of Nuclear Energy Advanced Reactor Technology program is in the pro-cess of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Benchmarks based on empirical data gathered during operation of the Fast Flux Test Facility (FFTF) as well as design documents and post-irradiation examination will aid in the validation of these software packages and the models and calculations they produce. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs« less

Advanced Demonstration and Test Reactor Options Study

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

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

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

A microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1985-02-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. To improve the analytical extrapolation of test results to full-size assembly bundles, the facility upgrade will increase the maximum size of the test bundle from 7 to 37 fuel pins. By creating a core convertor zone around the test location, the neutron spectrum incident on the test assembly will be hardened and the maximum energy deposited in the sample will be increased. In addition, a programmable Automated Reactor Control System (ARCS) willmore » permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations. A quantitative reliability analysis of the RTS shows that the unreliability, that is, the probability of failure, is acceptable for a 10 hour mission time or risk interval.« less

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

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

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

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

Experimental and numerical simulation of passive decay heat removal by sump cooling after cool melt down

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

Knebel, J.U.; Kuhn, D.; Mueller, U.

1997-12-01

This article presents the basic physical phenomena and scaling criteria of passive decay heat removal from a large coolant pool by single-phase and two-phase natural circulation. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and themore » Siemens AG. The article gives results of temperature and velocity measurements in the 1:20 linearly scaled SUCOS-2D test facility. The experiments are backed up by numerical calculations using the commercial software package Fluent. Finally, using the similarity analysis from above, the experimental results of the model geometry are scaled-up to the conditions in the prototype, allowing a first statement with regard to the feasibility of the sump cooling concept. 11 refs., 9 figs., 3 tabs.« less

Experiment for search for sterile neutrino at SM-3 reactor

NASA Astrophysics Data System (ADS)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

Zirconium Hydride Space Power Reactor design.

NASA Technical Reports Server (NTRS)

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

1972-01-01

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

Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

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

Rule, K.; Scott, J.; Larson, S.

1995-12-31

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methodsmore » for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.« less

Application of a fluidized bed reactor charged with aragonite for control of alkalinity, pH and carbon dioxide in marine recirculating aquaculture systems

USGS Publications Warehouse

Paul S Wills, PhD; Pfeiffer, Timothy; Baptiste, Richard; Watten, Barnaby J.

2016-01-01

Control of alkalinity, dissolved carbon dioxide (dCO2), and pH are critical in marine recirculating aquaculture systems (RAS) in order to maintain health and maximize growth. A small-scale prototype aragonite sand filled fluidized bed reactor was tested under varying conditions of alkalinity and dCO2 to develop and model the response of dCO2 across the reactor. A large-scale reactor was then incorporated into an operating marine recirculating aquaculture system to observe the reactor as the system moved toward equilibrium. The relationship between alkalinity dCO2, and pH across the reactor are described by multiple regression equations. The change in dCO2 across the small-scale reactor indicated a strong likelihood that an equilibrium alkalinity would be maintained by using a fluidized bed aragonite reactor. The large-scale reactor verified this observation and established equilibrium at an alkalinity of approximately 135 mg/L as CaCO3, dCO2 of 9 mg/L, and a pH of 7.0 within 4 days that was stable during a 14 day test period. The fluidized bed aragonite reactor has the potential to simplify alkalinity and pH control, and aid in dCO2 control in RAS design and operation. Aragonite sand, purchased in bulk, is less expensive than sodium bicarbonate and could reduce overall operating production costs.

Initial Back-to-Back Fission Chamber Testing in ATRC

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

Benjamin Chase; Troy Unruh; Joy Rempe

2014-06-01

Development and testing of in-pile, real-time neutron sensors for use in Materials Test Reactor experiments is an ongoing project at Idaho National Laboratory. The Advanced Test Reactor National Scientific User Facility has sponsored a series of projects to evaluate neutron detector options in the Advanced Test Reactor Critical Facility (ATRC). Special hardware was designed and fabricated to enable testing of the detectors in the ATRC. Initial testing of Self-Powered Neutron Detectors and miniature fission chambers produced promising results. Follow-on testing required more experiment hardware to be developed. The follow-on testing used a Back-to-Back fission chamber with the intent to providemore » calibration data, and a means of measuring spectral indices. As indicated within this document, this is the first time in decades that BTB fission chambers have been used in INL facilities. Results from these fission chamber measurements provide a baseline reference for future measurements with Back-to-Back fission chambers.« less

Directions of Small Group Research for the 1980's.

ERIC Educational Resources Information Center

Becker, Samuel L.

Speech communicators need to return to applied research, developing theoretical statements about how groups might work better and testing the validity of those statements. Among the tasks that group communication researchers might take up are determining the range of applicability of generalizations concerning different kinds of groups, studying…

76 FR 38680 - Notice of Availability of the Draft Environmental Impact Statement for the West Chocolate...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-01

... solar, wind and geothermal testing and development in a renewable energy evaluation area located near... pending geothermal lease application within the renewable energy evaluation area. This CDCA Draft Plan... Availability of the Draft Environmental Impact Statement for the West Chocolate Mountains Renewable Energy...

76 FR 34066 - Notice of Intent To Prepare an Environmental Impact Statement/Legislative Environmental Impact...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-10

.../Legislative Environmental Impact Statement (EIS/LEIS) to evaluate the potential environmental effects... allow the DoN to continue defense-related research, development, test and evaluation (RDT&E) and... high hazard training area for aerial gunnery, rocketry, electronic warfare and countermeasures, and...

Simultaneous and Successive Processes and K-ABC.

ERIC Educational Resources Information Center

Das, J. P.

1984-01-01

The article mentions six basic statements about sequential and simultaneous processes which are derived from A. Luria's clinical research. The Kaufman Assessment Battery for Children is then judged in terms of these statements. Suggestions for constructing tests which will entail planning as well as simultaneous and successive measures are…

Lunar Regolith Simulant Feed System for a Hydrogen Reduction Reactor System

NASA Technical Reports Server (NTRS)

Mueller, R. P.; Townsend, Ivan I., III

2009-01-01

One of the goals of In-Situ Resource Utilization (ISRU) on the moon is to produce oxygen from the lunar regolith which is present in the form of Ilmenite (FeTi03) and other compounds. A reliable and attainable method of extracting some of the oxygen from the lunar regolith is to use the hydrogen reduction process in a hot reactor to create water vapor which is then condensed and electrolyzed to obtain oxygen for use as a consumable. One challenge for a production system is to reliably acquire the regolith with an excavator hauler mobility platform and then introduce it into the reactor inlet tube which is raised from the surface and above the reactor itself. After the reaction, the hot regolith (-1000 C) must be expelled from the reactor for disposal by the excavator hauler mobility system. In addition, the reactor regolith inlet and outlet tubes must be sealed by valves during the reaction in order to allow collection of the water vapor by the chemical processing sub-system. These valves must be able to handle abrasive regolith passing through them as well as the heat conduction from the hot reactor. In 2008, NASA has designed and field tested a hydrogen reduction system called ROxygen in order to demonstrate the feasibility of extracting oxygen from lunar regolith. The field test was performed with volcanic ash known as Tephra on Mauna Kea volcano on the Big Island of Hawai'i. The tephra has similar properties to lunar regolith, so that it is regarded as a good simulant for the hydrogen reduction process. This paper will discuss the design, fabrication, operation, test results and lessons learned with the ROxygen regolith feed system as tested on Mauna Kea in November 2008.

Analysis and Down Select of Flow Passages for Thermal Hydraulic Testing of a SNAP Derived Reactor

NASA Technical Reports Server (NTRS)

Godfroy, T. J.; Sadasivan, P.; Masterson, S.

2007-01-01

As past of the Vision for Space Exploration, man will return to the moon. To enable safe and productive time on the lunar surface will require adequate power resources. To provide the needed power and to give mission planners all landing site possibilities, including a permanently dark crater, a nuclear reactor provides the most options. Designed to be l00kWt providing approx. 25kWe this power plants would be very effective in delivering dependable, site non-specific power to crews or robotic missions on the lunar surface. An affordable reference reactor based upon the successful SNAP program of the 1960's and early 1970's has been designed by Los Alamos National Laboratory that will meet such a requirement. Considering current funding, environmental, and schedule limitations this lunar surface power reactor will be tested using non-nuclear simulators to simulate the heat from fission reactions. Currently a 25kWe surface power SNAP derivative reactor is in the early process of design and testing with collaboration between Los Alamos National Laboratory, Idaho National Laboratory, Glenn Research Center, Marshall Space Flight Center, and Sandia National Laboratory to ensure that this new design is affordable and can be tested using non-nuclear methods as have proven so effective in the past. This paper will discuss the study and down selection of a flow passage concept for a approx. 25kWe lunar surface power reactor. Several different flow passages designs were evaluated using computational fluid dynamics to determine pressure drop and a structural assessment to consider thermal and stress of the passage walls. The reactor design basis conditions are discussed followed by passage problem setup and results for each concept. A recommendation for passage design is made with rationale for selection.

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .

Risk factors for bovine tuberculosis in low incidence regions related to the movements of cattle

PubMed Central

2013-01-01

Background Bovine tuberculosis (bTB) remains difficult to eradicate from low incidence regions partly due to the imperfect sensitivity and specificity of routine intradermal tuberculin testing. Herds with unconfirmed reactors that are incorrectly classified as bTB-negative may be at risk of spreading disease, while those that are incorrectly classified as bTB-positive may be subject to costly disease eradication measures. This analysis used data from Scotland in the period leading to Officially Tuberculosis Free recognition (1) to investigate the risks associated with the movements of cattle from herds with different bTB risk classifications and (2) to identify herd demographic characteristics that may aid in the interpretation of tuberculin testing results. Results From 2002 to 2009, for every herd with confirmed bTB positive cattle identified through routine herd testing, there was an average of 2.8 herds with at least one unconfirmed positive reactor and 18.9 herds with unconfirmed inconclusive reactors. Approximately 75% of confirmed bTB positive herds were detected through cattle with no known movements outside Scotland. At the animal level, cattle that were purchased from Scottish herds with unconfirmed positive reactors and a recent history importing cattle from endemic bTB regions were significantly more likely to react positively on routine intradermal tuberculin tests, while cattle purchased from Scottish herds with unconfirmed inconclusive reactors were significantly more likely to react inconclusively. Case-case comparisons revealed few demographic differences between herds with confirmed positive, unconfirmed positive, and unconfirmed inconclusive reactors, which highlights the difficulty in determining the true disease status of herds with unconfirmed tuberculin reactors. Overall, the risk of identifying reactors through routine surveillance decreased significantly over time, which may be partly attributable to changes in movement testing regulations and the volume of cattle imported from endemic regions. Conclusions Although the most likely source of bTB infections in Scotland was cattle previously imported from endemic regions, we found indirect evidence of transmission within Scottish cattle farms and cannot rule out the possibility of low level transmission between farms. Further investigation is needed to determine whether targeting herds with unconfirmed reactors and a history of importing cattle from high risk regions would benefit control efforts. PMID:24206865

REVIEW OF POWER AND HEAT REACTOR DESIGNS. Domestic and Foreign

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

Appleby, E.R., comp

1963-10-01

Unclassified information from domestic and foreign literature from January 1952 through September 1963 is compiled. Design characteristics and current information on the status of the individual designs are given, along with references for the associated literature. SNAP systems, proposed reactors, and chemonuclear and test reactors with characteristics similar to power reactors are included. The designs are indexed by name, location, type, and some special characteristics. (D.C.W.)

Small reactor power system for space application

NASA Technical Reports Server (NTRS)

Shirbacheh, M.

1987-01-01

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

Susceptibility constants of airborne bacteria to dielectric barrier discharge for antibacterial performance evaluation.

PubMed

Park, Chul Woo; Hwang, Jungho

2013-01-15

Dielectric barrier discharge (DBD) is a promising method to remove contaminant bioaerosols. The collection efficiency of a DBD reactor is an important factor for determining a reactor's removal efficiency. Without considering collection, simply defining the inactivation efficiency based on colony counting numbers for DBD as on and off may lead to overestimation of the inactivation efficiency of the DBD reactor. One-pass removal tests of bioaerosols were carried out to deduce the inactivation efficiency of the DBD reactor using both aerosol- and colony-counting methods. Our DBD reactor showed good performance for removing test bioaerosols for an applied voltage of 7.5 kV and a residence time of 0.24s, with η(CFU), η(Number), and η(Inactivation) values of 94%, 64%, and 83%, respectively. Additionally, we introduce the susceptibility constant of bioaerosols to DBD as a quantitative parameter for the performance evaluation of a DBD reactor. The modified susceptibility constant, which is the ratio of the susceptibility constant to the volume of the plasma reactor, has been successfully demonstrated for the performance evaluation of different sized DBD reactors under different DBD operating conditions. Our methodology will be used for design optimization, performance evaluation, and prediction of power consumption of DBD for industrial applications. Copyright © 2012 Elsevier B.V. All rights reserved.

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

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

Bohachek, Randolph Charles

2015-09-01

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

An underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

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

Boyack, B.E.

The PIUS reactor utilizes simplified, inherent, passive, or other innovative means to accomplish safety functions. Accordingly, the PIUS reactor is subject to the requirements of 10CFR52.47(b)(2)(i)(A). This regulation requires that the applicant adequately demonstrate the performance of each safety feature, interdependent effects among the safety features, and a sufficient data base on the safety features of the design to assess the analytical tools used for safety analysis. Los Alamos has assessed the quality and completeness of the existing and planned data bases used by Asea Brown Boveri to validate its safety analysis codes and other relevant data bases. Only amore » limited data base of separate effect and integral tests exist at present. This data base is not adequate to fulfill the requirements of 10CFR52.47(b)(2)(i)(A). Asea Brown Boveri has stated that it plans to conduct more separate effect and integral test programs. If appropriately designed and conducted, these test programs have the potential to satisfy most of the data base requirements of 10CFR52.47(b)(2)(i)(A) and remedy most of the deficiencies of the currently existing combined data base. However, the most important physical processes in PIUS are related to reactor shutdown because the PIUS reactor does not contain rodded shutdown and control systems. For safety-related reactor shutdown, PIUS relies on negative reactivity insertions from the moderator temperature coefficient and from boron entering the core from the reactor pool. Asea Brown Boveri has neither developed a direct experimental data base for these important processes nor provided a rationale for indirect testing of these key PIUS processes. This is assessed as a significant shortcoming. In preparing the conclusions of this report, test documentation and results have been reviewed for only one integral test program, the small-scale integral tests conducted in the ATLE facility.« less

Pneumatic Regolith Transfer Systems for In Situ Resource Utilization

NASA Technical Reports Server (NTRS)

Mueller, R. P.; Townsend, I. I.; Mantovani, J. G.; Zacny, Kris A.; Craft, Jack

2010-01-01

This slide presentation reviews the testing of a pneumatic system for transfering regolith, to be used for In Situ Resource Utilization (ISRU). Using both the simulated microgravity of parabolic flight and ground testing, the tests demonstrated that lunar regolith can be conveyed pneumatically into a simulated ISRU oxygen production plant reactor. The ground testing also demonstrated that the regolith can be expelled from the ISRU reactor for disposal or for other resource processing.

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

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

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

ETR ELECTRICAL BUILDING, TRA648. FLOOR PLANS FOR FIRST FLOOR AND ...

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

ETR ELECTRICAL BUILDING, TRA-648. FLOOR PLANS FOR FIRST FLOOR AND BASEMENT. SECTIONS. KAISER ETR-5528-MTR-648-A-2, 12/1955. INL INDEX NO. 532-0648-00-486-101402, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. CABLE TUNNEL. ...

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

ETR BUILDING, TRA-642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. CABLE TUNNEL. CAMERA FACING SOUTH INTO ETR ELECTRICAL BUILDING (TRA-648). INL NEGATIVE NO. HD46-20-2. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR ELECTRICAL BUILDING, TRA648. ELEVATIONS AND DETAILS. ROOF PLAN. DOOR ...

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

ETR ELECTRICAL BUILDING, TRA-648. ELEVATIONS AND DETAILS. ROOF PLAN. DOOR SCHEDULE. KAISER ETR-5528-MTR-648-A-3, 1/1956. INL INDEX NO. 532-0648-00-486-101403, REV. 5. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. BASEMENT. ROLLUP DOOR TO CUBICLE POSTS ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. ROLLUP DOOR TO CUBICLE POSTS CAUTION SIGNS BECAUSE OF SODIUM HAZARD WITHIN. INL NEGATIVE NO. HD24-3-1. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-29

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

MTR WING, TRA604, INTERIOR. BASEMENT. WEST CORRIDOR. CAMERA FACES NORTH. ...

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

MTR WING, TRA-604, INTERIOR. BASEMENT. WEST CORRIDOR. CAMERA FACES NORTH. HVAC AREA IS AT RIGHT OF CORRIDOR. INL NEGATIVE NO. HD46-13-3. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR HEAT EXCHANGER BUILDING, TRA644. DETAIL OF SOUTH SIDE BUILDING ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. DETAIL OF SOUTH SIDE BUILDING INSET. DEMINERALIZER WING AT RIGHT. CAMERA FACING NORTH. INL NEGATIVE NO. HD46-36-2. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR STACK, TRA710, CONTEXTUAL VIEW, CAMERA FACING SOUTH. PERIMETER SECURITY ...

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

MTR STACK, TRA-710, CONTEXTUAL VIEW, CAMERA FACING SOUTH. PERIMETER SECURITY FENCE AND SECURITY LIGHTING IN VIEW AT LEFT. INL NEGATIVE NO. HD52-1-1. Mike Crane, Photographer, 5/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REACTIVITY MEASUREMENT FACILITY, UNDER CONSTRUCTION OVER MTR CANAL IN BASEMENT ...

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

REACTIVITY MEASUREMENT FACILITY, UNDER CONSTRUCTION OVER MTR CANAL IN BASEMENT OF MTR BUILDING, TRA-603. WOOD PLANKS REST ON CANAL WALL OBSERVABLE IN FOREGROUND. INL NEGATIVE NO. 11745. Unknown Photographer, 8/20/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

VIEW OF PROCESS DEVELOPMENT PILE (PDP) TANK, LOOKING WESTSOUTHWEST, BASEMENT ...

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

VIEW OF PROCESS DEVELOPMENT PILE (PDP) TANK, LOOKING WEST-SOUTHWEST, BASEMENT LEVEL -15’. EDGE O FRESONANCE TEST REACTOR (RTR), LATER KNOWN AS LATTICE TEST REACTOR (LTR), VISIBLE TO RIGHT OF PDP TANK - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

MTR WING, TRA604. PRECAST CONCRETE PANELS AND DIMENSIONS FOR PANELS ...

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

MTR WING, TRA-604. PRECAST CONCRETE PANELS AND DIMENSIONS FOR PANELS K THROUGH Q. BLAW-KNOX 3150-804-21, SHEET #2, 11/1950. INL INDEX NO. 531-0604-62-098-100645, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

NORTH BASEMENT WALL. IBEAM COLUMNS HAVE BEEN ENCASED IN CONCRETE. ...

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

NORTH BASEMENT WALL. I-BEAM COLUMNS HAVE BEEN ENCASED IN CONCRETE. STEEL BEAMS LAY ACROSS FIRST FLOOR AWAITING CONCRETE POUR. CAMERA LOOKS SOUTHWEST. INL NEGATIVE NO. 735. Unknown Photographer, 10/6/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. ONE OF THREE EVAPORATORS BEFORE IT ...

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

PROCESS WATER BUILDING, TRA-605. ONE OF THREE EVAPORATORS BEFORE IT IS INSTALLED IN UPPER LEVEL OF EAST HALF OF BUILDING. INL NEGATIVE NO. 1533. Unknown Photographer, 3/1/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. BENCH MARK AND ELEVATION LOCATIONS, FLOOR LOADING DATA, ...

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

ETR, TRA-642. BENCH MARK AND ELEVATION LOCATIONS, FLOOR LOADING DATA, CRANE WORKING AREAS. PHILLIPS PETROLEUM COMPANY ETR-D-1584, 5/1959. INL INDEX NO. 532-0642-00-706-020323, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

75 FR 27368 - Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-14

...., Aerotest Radiography and Research Reactor; Notice of Consideration of Approval of Transfer and Conforming Amendment, Opportunity for a Hearing, and Order Imposing Procedures for Access to Sensitive Unclassified Non... Manager, Research and Test Reactors Licensing Branch, Division of Policy and Rulemaking, Office of Nuclear...

10 CFR 100.11 - Determination of exclusion area, low population zone, and population center distance.

Code of Federal Regulations, 2011 CFR

2011-01-01

... population center distance. 100.11 Section 100.11 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactors § 100.11 Determination of exclusion area, low population zone, and population...

10 CFR 100.11 - Determination of exclusion area, low population zone, and population center distance.

Code of Federal Regulations, 2010 CFR

2010-01-01

... population center distance. 100.11 Section 100.11 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactors § 100.11 Determination of exclusion area, low population zone, and population...

77 FR 55877 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-11

...-492- 3668; email: [email protected] . NRC's Agencywide Documents Access and Management System... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance... Systems for Boiling Water Reactor Power Plants.'' This regulatory guide is being revised to: (1) Expand...

Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

NASA Technical Reports Server (NTRS)

Martin, James J.; Reid, Robert S.

2004-01-01

A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

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

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

None

1960-10-31

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

Detection, evaluation and treatment of diabetes mellitus in chronic pancreatitis: recommendations from PancreasFest 2012.

PubMed

Rickels, Michael R; Bellin, Melena; Toledo, Frederico G S; Robertson, R Paul; Andersen, Dana K; Chari, Suresh T; Brand, Randall; Frulloni, Luca; Anderson, Michelle A; Whitcomb, David C

2013-01-01

Diabetes and glucose intolerance are common complications of chronic pancreatitis, yet clinical guidance on their detection, classification, and management is lacking. A working group reviewed the medical problems, diagnostic methods, and treatment options for chronic pancreatitis-associated diabetes for a consensus meeting at PancreasFest 2012. Guidance Statement 1.1: Diabetes mellitus is common in chronic pancreatitis. While any patient with chronic pancreatitis should be monitored for development of diabetes, those with long-standing duration of disease, prior partial pancreatectomy, and early onset of calcific disease may be at higher risk. Those patients developing diabetes mellitus are likely to have co-existing pancreatic exocrine insufficiency. Guidance Statement 1.2: Diabetes occurring secondary to chronic pancreatitis should be recognized as pancreatogenic diabetes (type 3c diabetes). Guidance Statement 2.1: The initial evaluation should include fasting glucose and HbA1c. These tests should be repeated annually. Impairment in either fasting glucose or HbA1c requires further evaluation. Guidance Statement 2.2: Impairment in either fasting glucose or HbA1c should be further evaluated by a standard 75 g oral glucose tolerance test. Guidance Statement 2.3: An absent pancreatic polypeptide response to mixed-nutrient ingestion is a specific indicator of type 3c diabetes. Guidance Statement 2.4: Assessment of pancreatic endocrine reserve, and importantly that of functional beta-cell mass, should be performed as part of the evaluation and follow-up for total pancreatectomy with islet autotransplantation (TPIAT). Guidance Statement 3: Patients with pancreatic diabetes shall be treated with specifically tailored medical nutrition and pharmacologic therapies. Physicians should evaluate and treat glucose intolerance in patients with pancreatitis. Copyright © 2013 IAP and EPC. All rights reserved.

Detection, Evaluation and Treatment of Diabetes Mellitus in Chronic Pancreatitis: Recommendations from PancreasFest 2012

PubMed Central

Rickels, Michael R.; Bellin, Melena; Toledo, Frederico G.S.; Robertson, R. Paul; Andersen, Dana K.; Chari, Suresh T.; Brand, Randall; Frulloni, Luca; Anderson, Michelle A.; Whitcomb, David C.

2013-01-01

Description Diabetes and glucose intolerance are common complications of chronic pancreatitis, yet clinical guidance on their detection, classification, and management is lacking. Methods A working group reviewed the medical problems, diagnostic methods, and treatment options for chronic pancreatitis-associated diabetes for a consensus meeting at PancreasFest 2012. Results Guidance Statement 1.1 Diabetes mellitus is common in chronic pancreatitis. While any patient with chronic pancreatitis should be monitored for development of diabetes, those with long-standing duration of disease, prior partial pancreatectomy, and early onset of calcific disease may be at higher risk. Those patients developing diabetes mellitus are likely to have co-existing pancreatic exocrine insufficiency. Guidance Statement 1.2 Diabetes occurring secondary to chronic pancreatitis should be recognized as pancreatogenic diabetes (type 3c diabetes). Guidance Statement 2.1 The initial evaluation should include fasting glucose and HbA1c. These tests should be repeated annually. Impairment in either fasting glucose or HbA1c requires further evaluation. Guidance Statement 2.2 Impairment in either fasting glucose or HbA1c should be further evaluated by a standard 75 gram oral glucose tolerance test. Guidance Statement 2.3 An absent pancreatic polypeptide response to mixed-nutrient ingestion is a specific indicator of type 3c diabetes. Guidance Statement 2.4 Assessment of pancreatic endocrine reserve, and importantly that of functional beta-cell mass, should be performed as part of the evaluation and follow-up for total pancreatectomy with islet autotransplantation (TPIAT). Guidance Statement 3 Patients with pancreatic diabetes shall be treated with specifically tailored medical nutrition and pharmacologic therapies. Conclusions Physicians should evaluate and treat glucose intolerance in patients with pancreatitis. PMID:23890130

CUTE: A Concolic Unit Testing Engine for C

DTIC Science & Technology

2005-01-01

We also introduce program units of a simple C-like language (cf. [20]). We present how CUTE instruments programs and performs concolic execution. We...works for a simple C-like language shown in Figure 2. START represents the first statement of a program under test. Each statement has an optional...is a variable, c is a constant p ::= v = v | v 6= v | v < v | v ≤ v | v ≥ v | v > v Figure 2: Syntax of a simple C-like language the inputs at the

Consensus Statement of Society of Abdominal Radiology Disease-Focused Panel on Barium Esophagography in Gastroesophageal Reflux Disease.

PubMed

Levine, Marc S; Carucci, Laura R; DiSantis, David J; Einstein, David M; Hawn, Mary T; Martin-Harris, Bonnie; Katzka, David A; Morgan, Desiree E; Rubesin, Stephen E; Scholz, Francis J; Turner, Mary Ann; Wolf, Ellen L; Canon, Cheri L

2016-11-01

The Society of Abdominal Radiology established a panel to prepare a consensus statement on the role of barium esophagography in gastroesophageal reflux disease (GERD), as well as recommended techniques for performing the fluoroscopic examination and the gamut of findings associated with this condition. Because it is an inexpensive, noninvasive, and widely available study that requires no sedation, barium esophagography may be performed as the initial test for GERD or in conjunction with other tests such as endoscopy.

Comparative effectiveness of allergy testing method in driving immunotherapy outcomes.

PubMed

Kaffenberger, Thomas M; Dedhia, Raj C; Schwarzbach, Hannah L; Mady, Leila J; Lee, Stella E

2018-05-01

Skin-prick testing (SPT), in vitro testing (IVT), and intradermal-dilutional testing (IDT) are methods to detect patient sensitivities to specific allergens and direct immunotherapy dosing. We used objective and subjective measures of improvement to compare outcomes based on test method. Patients underwent 1 of 3 protocols: SPT, screening SPT followed by IDT, or IVT. We used institution billing data to do a cost analysis of these tests. The time to maintenance (TTM) therapy was analyzed and patients were stratified into high and low reactors. The Rhinoconjunctivitis Quality of Life Questionnaire (RQLQ) was used to quantify symptoms pre-maintenance and post-maintenance. Of 177 patients (SPT, n = 40; IVT, n = 91; IDT, n = 46), 115 (SPT, n = 35; IVT, n = 39; IDT, n = 41) were high reactors. Out of 90 patients (SPT, n = 17; IVT, n = 37; IDT, n = 36) reaching maintenance, 58 were high reactors (SPT, n = 15; IVT, n = 12; IDT, n = 31). Overall, SPT, IVT, and IDT median TTM were 542, 329, and 578.5 days, respectively. IDT TTM was shorter compared to IVT overall and in high reactors (hazard ratio [HR] = 1.91, p = 0.02; HR = 2.12, p = 0.03), but was not significant compared to SPT high reactors (p = 0.33). The IDT cost was $62.66, translating to an incremental cost-effectiveness ratio of $0.23 per day of shortened TTM. Median RQLQ change for the SPT, IVT, and IDT groups was 6.5, 1, and 1.5, respectively, but was not significant (p = 0.60). IDT reached maintenance immunotherapy quicker than IVT but there was no difference compared to SPT. TTM did not correlate with improvements in patient symptoms between testing methods. This study represents a novel comparison of outcomes based on initial allergy testing method. © 2017 ARS-AAOA, LLC.

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

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

John Darrell Bess

2008-06-01

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

Flow reversal power limit for the HFBR

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

Cheng, L.Y.; Tichler, P.R.

The High Flux Beam Reactor (HFBR) is a pressurized heavy water moderated and cooled research reactor that began operation at 40 MW. The reactor was subsequently upgraded to 60 MW and operated at that level for several years. The reactor undergoes a buoyancy-driven reversal of flow in the reactor core following certain postulated accidents. Questions which were raised about the afterheat removal capability during the flow reversal transition led to a reactor shutdown and subsequent resumption of operation at a reduced power of 30 MW. An experimental and analytical program to address these questions is described in this report. Themore » experiments were single channel flow reversal tests under a range of conditions. The analytical phase involved simulations of the tests to benchmark the physical models and development of a criterion for dryout. The criterion is then used in simulations of reactor accidents to determine a safe operating power level. It is concluded that the limit on the HFBR operating power with respect to the issue of flow reversal is in excess of 60 MW. Direct use of the experimental results and an understanding of the governing phenomenology supports this conclusion.« less

Development of a small specimen test machine to evaluate irradiation embrittlement of fusion reactor materials

NASA Astrophysics Data System (ADS)

Ishii, T.; Ohmi, M.; Saito, J.; Hoshiya, T.; Ooka, N.; Jitsukawa, S.; Eto, M.

2000-12-01

Small specimen test techniques (SSTT) are essential to use an accelerator-driven deuterium-lithium stripping reaction neutron source for the study of fusion reactor materials because of the limitation of the available irradiation volume. A remote-controlled small punch (SP) test machine was developed at the hot laboratory of the Japan Materials Testing Reactor (JMTR) in the Japan Atomic Energy Research Institute (JAERI). This report describes the SP test method and machine for use in a hot cell, and test results on irradiated ferritic steels. The specimen was either a coupon 10×10×0.25 mm 3 or a TEM disk 3 mm in diameter by 0.25 mm in thickness. Tests can be performed at temperatures ranging from 93 to 1123 K in a vacuum or in an inert gas environment. The ductile to brittle transition temperature of the irradiated ferritic steel as determined by the SP test is also evaluated.

Fiber Attachment Module Experiment (FAME): Using a Multiplexed Miniature Hollow Fiber Membrane Bioreactor Solution for Rapid Process Testing

NASA Astrophysics Data System (ADS)

Lunn, Griffin; Wheeler, Raymond; Hummerick, Mary; Birmele, Michele; Richards, Jeffrey; Coutts, Janelle; Koss, Lawrence; Spencer, Lashelle.; Johnsey, Marissa; Ellis, Ronald

Bioreactor research, even today, is mostly limited to continuous stirred-tank reactors (CSTRs). These are not an option for microgravity applications due to the lack of a gravity gradient to drive aeration as described by the Archimedes principle. This has led to testing of Hollow Fiber Membrane Bioreactors (HFMBs) for microgravity applications, including possible use for wastewater treatment systems for the International Space Station (ISS). Bioreactors and filtration systems for treating wastewater could avoid the need for harsh pretreatment chemicals and improve overall water recovery. However, the construction of these reactors is difficult and commercial off-the-shelf (COTS) versions do not exist in small sizes. We have used 1-L modular HFMBs in the past, but the need to perform rapid testing has led us to consider even smaller systems. To address this, we designed and built 125-mL, rectangular reactors, which we have called the Fiber Attachment Module Experiment (FAME) system. A polycarbonate rack of four square modules was developed with each module containing removable hollow fibers. Each FAME reactor is self-contained and can be easily plumbed with peristaltic and syringe pumps for continuous recycling of fluids and feeding, as well as fitted with sensors for monitoring pH, dissolved oxygen, and gas measurements similar to their larger counterparts. The first application tested in the FAME racks allowed analysis of over a dozen fiber surface treatments and three inoculation sources to achieve rapid reactor startup and biofilm attachment (based on carbon oxidation and nitrification of wastewater). With these miniature FAME reactors, data for this multi-factorial test were collected in duplicate over a six-month period; this greatly compressed time period required for gathering data needed to study and improve bioreactor performance.

PROCESS WATER BUILDING, TRA605. AERIAL TAKEN WHILE SEVERAL PIPE TRENCHES ...

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

PROCESS WATER BUILDING, TRA-605. AERIAL TAKEN WHILE SEVERAL PIPE TRENCHES REMAINED OPEN. CAMERA FACES EASTERLY. NOTE DUAL PIPES BETWEEN REACTOR BUILDING AND NORTH SIDE OF PROCESS WATER BUILDING. PIPING NEAR WORKING RESERVOIR HEADS FOR RETENTION RESERVOIR. PIPE FROM DEMINERALIZER ENTERS MTR FROM NORTH. SEE ALSO TRENCH FOR COOLANT AIR DUCT AT SOUTH SIDE OF MTR AND LEADING TO FAN HOUSE AND STACK. INL NEGATIVE NO. 2966-A. Unknown Photographer, 7/31/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

An improved external recycle reactor for determining gas-solid reaction kinetics

NASA Technical Reports Server (NTRS)

Miller, Irvin M.; Hoyt, Ronald F.

1987-01-01

These improvements in the recycle system effectively eliminate initial concentration variation by two modifications: (1) a vacuum line connection to the recycle loop which permits this loop to be evacuated and then filled with the test gas mixture to slightly above atmospheric pressure; and (2) a bypass line across the reactor which permits the reactor to be held under vacuum while the rest of the recycle loop is filled with test gas. A three-step procedure for bringing the feed gas mixture into contact with the catalyst at time zero is described.

ETR CRITICAL FACILITY, TRA654. CONTEXTUAL VIEW. CAMERA ON ROOF OF ...

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

ETR CRITICAL FACILITY, TRA-654. CONTEXTUAL VIEW. CAMERA ON ROOF OF MTR BUILDING AND FACING SOUTH. ETR AND ITS COOLANT BUILDING AT UPPER PART OF VIEW. ETR COOLING TOWER NEAR TOP EDGE OF VIEW. EXCAVATION AT CENTER IS FOR ETR CF. CENTER OF WHICH WILL CONTAIN POOL FOR REACTOR. NOTE CHOPPER TUBE PROCEEDING FROM MTR IN LOWER LEFT OF VIEW, DIAGONAL TOWARD LEFT. INL NEGATIVE NO. 56-4227. Jack L. Anderson, Photographer, 12/18/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Pipe connector

DOEpatents

Sullivan, Thomas E.; Pardini, John A.

1978-01-01

A safety test facility for testing sodium-cooled nuclear reactor components includes a reactor vessel and a heat exchanger submerged in sodium in the tank. The reactor vessel and heat exchanger are connected by an expansion/deflection pipe coupling comprising a pair of coaxially and slidably engaged tubular elements having radially enlarged opposed end portions of which at least a part is of spherical contour adapted to engage conical sockets in the ends of pipes leading out of the reactor vessel and in to the heat exchanger. A spring surrounding the pipe coupling urges the end portions apart and into engagement with the spherical sockets. Since the pipe coupling is submerged in liquid a limited amount of leakage of sodium from the pipe can be tolerated.

CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

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

Pytel, K.; Mieleszczenko, W.; Lechniak, J.

2010-03-01

The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

Catalytic Reactor for Inerting of Aircraft Fuel Tanks

DTIC Science & Technology

1974-06-01

Aluminum Panels After Triphase Corrosion Test 79 35 Inerting System Flows in Various Flight Modes 82 36 High Flow Reactor Parametric Data 84 37 System...AD/A-000 939 CATALYTIC REACTOR FOR INERTING OF AIRCRAFT FUEL TANKS George H. McDonald, et al AiResearch Manufacturing Company Prepared for: Air Force...190th Street 2b. GROUP Torrance, California .. REPORT TITLE CATALYTIC REACTOR FOR INERTING OF AIRCRAFT FUEL TANKS . OESCRIP TIVE NOTEs (Thpe of refpoft

Development of Inspection and Repair Technology for Heat Exchanger Tubes in Fast Breeder Reactors

DTIC Science & Technology

2009-06-01

Technology for Heat Exchanger Tubes in Fast Breeder Reactors Akihiko NISHIMURA *1 , Takahisa SHOBU, Kiyoshi OKA, Toshihiko YAMAGUCHI, Yukihiro SHIMADA...fast breeder reactors (FBRs). It comprises a laser processing head combined with an eddy current testing unit. Ultrashort laser pulse ablation is used...be applied in the main- tenance of large structures such as nuclear reactors and chemical factories [1]. Internal access to a blanket cooling pipe

Preliminary neutronic analysis of a cavity test reactor

NASA Technical Reports Server (NTRS)

Whitmarsh, C. L., Jr.

1973-01-01

A reference configuration was calculated for a cavity test reactor to be used for testing the gascore nuclear rocket concept. A thermal flux of 4.1 x 10 to the 14th power neutrons per square centimeter per second in the cavity was provided by a driver fuel loading of 6.4 kg of enriched uranium in MTR fuel elements. The reactor was moderated and cooled by heavy water and reflected with 25.4 cm of beryllium. Power generation of 41.3 MW in the driver fuel is rejected to a heat sink. Design effort was directed toward minimization of driver power while maintaining 2.7 MW in the cavity during a test run. Ancillary data on material reactivity worths, reactivity coefficients, flux spectra, and power distributions are reported.

High-intensity power-resolved radiation imaging of an operational nuclear reactor.

PubMed

Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

2015-10-09

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

High-intensity power-resolved radiation imaging of an operational nuclear reactor

PubMed Central

Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

2015-01-01

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

Identifying and clarifying values and reason statements that promote effective food parenting practices, using intensive interviews

USDA-ARS?s Scientific Manuscript database

The objective was to generate and test parents' understanding of values and associated reason statements to encourage effective food parenting practices. This study was cross-sectional. Sixteen parents from different ethnic groups (African American, white, and Hispanic) living with their 3- to 5-yea...

77 FR 71446 - Notice of Availability of the Final Environmental Impact Statement for the West Chocolate...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-30

... geothermal energy testing and development. The Final EIS also analyzes the potential environmental impacts of... facilitate appropriate development of geothermal, solar, and wind energy in the REEA and make land use plan... Availability of the Final Environmental Impact Statement for the West Chocolate Mountains Renewable Energy...

27 CFR 7.29 - Prohibited practices.

Code of Federal Regulations, 2011 CFR

2011-04-01

... the words “strong,” “full strength,” “extra strength,” “high test,” “high proof,” “pre-war strength... of a competitor's products. (3) Any statement, design, device, or representation which is obscene or indecent. (4) Any statement, design, device, or representation of or relating to analyses, standards, or...

27 CFR 7.29 - Prohibited practices.

Code of Federal Regulations, 2012 CFR

2012-04-01

... the words “strong,” “full strength,” “extra strength,” “high test,” “high proof,” “pre-war strength... of a competitor's products. (3) Any statement, design, device, or representation which is obscene or indecent. (4) Any statement, design, device, or representation of or relating to analyses, standards, or...

27 CFR 7.29 - Prohibited practices.

Code of Federal Regulations, 2014 CFR

2014-04-01

... the words “strong,” “full strength,” “extra strength,” “high test,” “high proof,” “pre-war strength... of a competitor's products. (3) Any statement, design, device, or representation which is obscene or indecent. (4) Any statement, design, device, or representation of or relating to analyses, standards, or...

The Effect of Three Message Organization Variables Upon Listener Comprehension.

ERIC Educational Resources Information Center

Johnson, Arlee W.

Public speaking texts urge speakers to organize their message in order to increase their audience's comprehension of it. Tests were run to determine if listeners understand better when three message organization variables are employed in a speech: explicit statement of the central idea, explicit statement of the main points, and transitions before…

Beyond Test Prep: Moving the Mission Statement into the Classroom

ERIC Educational Resources Information Center

Jones, Emily

2010-01-01

A random walk through the mission statements of independent schools shows an admirable determination to educate students for an unknowable future, for creativity and problem solving, for responsible citizenship, for resiliency. Nevertheless, many of these same schools are constrained to work towards their mission-central goals in time stolen from…

29 CFR 4231.8 - Notice of merger or transfer.

Code of Federal Regulations, 2012 CFR

2012-07-01

... it is a de minimis transaction as defined in § 4231.7 (and, if so, including an enrolled actuary's... that exists after the transaction, one of the following statements, certified by an enrolled actuary..., indicating which is the applicable test. (ii) A statement of the basis on which the actuary has determined...

29 CFR 4231.8 - Notice of merger or transfer.

Code of Federal Regulations, 2011 CFR

2011-07-01

... it is a de minimis transaction as defined in § 4231.7 (and, if so, including an enrolled actuary's... that exists after the transaction, one of the following statements, certified by an enrolled actuary..., indicating which is the applicable test. (ii) A statement of the basis on which the actuary has determined...

29 CFR 4231.8 - Notice of merger or transfer.

Code of Federal Regulations, 2013 CFR

2013-07-01

... it is a de minimis transaction as defined in § 4231.7 (and, if so, including an enrolled actuary's... that exists after the transaction, one of the following statements, certified by an enrolled actuary..., indicating which is the applicable test. (ii) A statement of the basis on which the actuary has determined...

40 CFR 280.94 - Allowable mechanisms and combinations of mechanisms.

Code of Federal Regulations, 2010 CFR

2010-07-01

... requirements of the financial test under this rule, the financial statements of the owner or operator are not consolidated with the financial statements of the guarantor. [53 FR 43370, Oct. 26, 1988, as amended at 58 FR... OPERATORS OF UNDERGROUND STORAGE TANKS (UST) Financial Responsibility § 280.94 Allowable mechanisms and...

Novice Users' Misconceptions of BASIC Programming Statements. Report No. 82-1.

ERIC Educational Resources Information Center

Bayman, Piraye; Mayer, Richard E.

The nature of novice programmers' mental models for BASIC statements following preliminary BASIC instruction was assessed with 30 undergraduates who were taught BASIC through a self-paced, mastery manual and who were simultaneously given hands-on access to an Apple II microcomputer. Following instruction, the students were tested to determine…

Three Year Evaluations for Handicapped Students. Position Statement [and] Supporting Paper for Position Statement on Reevaluation.

ERIC Educational Resources Information Center

National Association of School Psychologists, Washington, DC.

The National Association of School Psychologists' position on 3-year evaluations for handicapped students calls for a flexible, individualized approach to student evaluations, rather than a standardized testing system. Specific issues to be addressed by these evaluations are discussed, including effectiveness of the student's individual education…

29 CFR 4231.8 - Notice of merger or transfer.

Code of Federal Regulations, 2010 CFR

2010-07-01

... it is a de minimis transaction as defined in § 4231.7 (and, if so, including an enrolled actuary's... that exists after the transaction, one of the following statements, certified by an enrolled actuary..., indicating which is the applicable test. (ii) A statement of the basis on which the actuary has determined...

12 CFR 620.11 - Content of quarterly report to shareholders.

Code of Federal Regulations, 2012 CFR

2012-01-01

... major income statement caption is less than 15 percent of average net income for the 3 most recent... the 3 most recent fiscal years, the average loss shall be used for purposes of this test. (3) The... would substantially duplicate the disclosure contained in the most recent audited financial statements...

76 FR 63668 - Guidelines for Preparing and Reviewing Licensing Applications for the Production of Radioisotopes

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-13

... Licensing of Non-Power Reactors: Format and Content,'' for the Production of Radioisotopes and NUREG-1537, part 2, ``Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors... production facility and the Research and Test Reactor Licensing Branch (PRLB) of the Division of Policy and...

Early Program Development

NASA Image and Video Library

1963-01-01

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

LOFT. Reactor apparatus leaves A&M building (TAN607). Shielded locomotive has ...

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

LOFT. Reactor apparatus leaves A&M building (TAN-607). Shielded locomotive has aerojet logo, which replaced old general electric logo, pulls reactor from assembly shop on dolly. Camera facing easterly. Date: 1973. INEEL negative no. 73-3700 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

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

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.

1994-04-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ``flooded cavity``, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array canmore » deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications.« less

CUTS FOR MTR EXCAVATION ILLUSTRATE SEDIMENTARY MANTLE OF SOIL AND ...

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

CUTS FOR MTR EXCAVATION ILLUSTRATE SEDIMENTARY MANTLE OF SOIL AND GRAVEL OVERLAYING LAVA ROCK FIFTY FEET BELOW. SAGEBRUSH HAS BEEN SCOURED FROM REST OF SITE. CAMERA PROBABLY FACES SOUTHWEST. INL NEGATIVE NO. 67. Unknown Photographer, 6/4/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605, INTERIOR. FIRST FLOOR. ELECTRICAL EQUIPMENT IN ...

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

PROCESS WATER BUILDING, TRA-605, INTERIOR. FIRST FLOOR. ELECTRICAL EQUIPMENT IN LEFT HALF OF VIEW. CAMERA IS IN NORTHWEST CORNER FACING SOUTHEAST. INL NEGATIVE NO. HD46-27-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. FIRST FLOOR. INSIDE UTILITY CORRIDOR ALONG ...

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

ETR BUILDING, TRA-642, INTERIOR. FIRST FLOOR. INSIDE UTILITY CORRIDOR ALONG SOUTH PERIMETER WALL (COMMON TO ELECTRICAL BUILDING, TRA-648). CAMERA FACES WEST. INL NEGATIVE NO. HD46-16-2. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID