Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification

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

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

2017-07-05

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

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

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

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

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

Renewing Liquid Fueled Molten Salt Reactor Research and Development

NASA Astrophysics Data System (ADS)

Towell, Rusty; NEXT Lab Team

2016-09-01

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

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

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

Travelli, A.; Technology Development

2003-01-01

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

Rover nuclear rocket engine program: Overview of rover engine tests

NASA Technical Reports Server (NTRS)

Finseth, J. L.

1991-01-01

The results of nuclear rocket development activities from the inception of the ROVER program in 1955 through the termination of activities on January 5, 1973 are summarized. This report discusses the nuclear reactor test configurations (non cold flow) along with the nuclear furnace demonstrated during this time frame. Included in the report are brief descriptions of the propulsion systems, test objectives, accomplishments, technical issues, and relevant test results for the various reactor tests. Additionally, this document is specifically aimed at reporting performance data and their relationship to fuel element development with little or no emphasis on other (important) items.

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.

TECHNICAL SCOPE OF GAS-COOLED REACTOR FUEL ELEMENT IRRADIATION PROGRAM

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

None

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

Sister Lab Program Prospective Partner Nuclear Profile: Indonesia

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

Bissani, M; Tyson, S

2006-12-14

Indonesia has participated in cooperative technical programs with the IAEA since 1957, and has cooperated with regional partners in all of the traditional areas where nuclear science is employed: in medicine, public health (such as insect control and eradication programs), agriculture (e.g. development of improved varieties of rice), and the gas and oil industries. Recently, Indonesia has contributed significantly to the Reduced Enrichment Research and Training Reactor (RERTR) Program by conducting experiments to confirm the feasibility of Mo-99 production using high-density low enriched uranium (LEU) fuel, a primary goal of the RERTR Program. Indonesia's first research reactor, the TRIGA Markmore » II at Bandung, began operation in 1964 at 250 kW and was subsequently upgraded in 1971 to 1 MW and further upgraded in 2000 to 2 MW. This reactor was joined by another TRIGA Mark II, the 100-kW Kartini-PPNY at Yogyakarta, in 1979, and by the 30-MW G.A. Siwabessy multipurpose reactor in Serpong, which achieved criticality in July 1983. A 10-MW radioisotope production reactor, to be called the RPI-10, also was proposed for construction at Serpong in the late 1990s, but the project apparently was not carried out. In the five decades since its nuclear research program began, Indonesia has trained a cadre of scientific and technical staff who not only operate and conduct research with the current facilities, but also represent the nucleus of a skilled labor pool to support development of a nuclear power program. Although Indonesia's previous on-again, off-again consideration of nuclear power has not gotten very far in the past, it now appears that Indonesia again is giving serious consideration to beginning a national nuclear energy program. In June 2006, Research and Technology Minister Kusmayanto Kadiman said that his ministry was currently putting the necessary procedures in place to speed up the project to acquire a nuclear power plant, indicating that, ''We will need around five years to complete the project. If we can start the study, go to tender, and sign the contract for the project this year, the power plant could be on stream by 2011''. While this ambitious schedule may be a bit unrealistic, it suggests new momentum to move forward on the project. The favored site for the proposed plant is the Muria Peninsula, located on Java's north central coast.« less

Current Abstracts Nuclear Reactors and Technology

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

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

1993-01-01

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

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

Code of Federal Regulations, 2012 CFR

2012-01-01

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

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

Code of Federal Regulations, 2013 CFR

2013-01-01

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

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

DOE PAGES

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

2016-10-31

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

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

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

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

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

ATF Neutron Irradiation Program Irradiation Vehicle Design Concepts

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

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

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

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

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

Ingersoll, Daniel T

2007-01-01

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

Goals of thermionic program for space power

NASA Technical Reports Server (NTRS)

English, R. E.

1981-01-01

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

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

Code of Federal Regulations, 2011 CFR

2011-01-01

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

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

Code of Federal Regulations, 2010 CFR

2010-01-01

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

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

Code of Federal Regulations, 2014 CFR

2014-01-01

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

The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

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

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

2009-05-01

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

Assessment of Sensor Technologies for Advanced Reactors

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

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

2016-10-01

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

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

NASA Technical Reports Server (NTRS)

Knight, J. C.

1996-01-01

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

Small and medium power reactors 1987

NASA Astrophysics Data System (ADS)

1987-12-01

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

Dry coal feeder development program at Ingersoll-Rand Research, Incorporated. [for coal gasification systems

NASA Technical Reports Server (NTRS)

Mistry, D. K.; Chen, T. N.

1977-01-01

A dry coal screw feeder for feeding coal into coal gasification reactors operating at pressures up to 1500 psig is described. Results on the feeder under several different modes of operation are presented. In addition, three piston feeder concepts and their technical and economical merits are discussed.

The U.S. RERTR program status and progress.

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

Travelli, A.

1998-01-21

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

Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

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

Okrent, D.

1997-06-23

This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, withmore » the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.« less

Reactor technology assessment and selection utilizing systems engineering approach

NASA Astrophysics Data System (ADS)

Zolkaffly, Muhammed Zulfakar; Han, Ki-In

2014-02-01

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

Autonomous Control Capabilities for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

2004-02-01

The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

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.)

Historical perspectives - The role of the NASA Lewis Research Center in the national space nuclear power programs

NASA Technical Reports Server (NTRS)

Bloomfield, H. S.; Sovie, R. J.

1991-01-01

The history of the NASA Lewis Research Center's role in space nuclear power programs is reviewed. Lewis has provided leadership in research, development, and the advancement of space power and propulsion systems. Lewis' pioneering efforts in nuclear reactor technology, shielding, high temperature materials, fluid dynamics, heat transfer, mechanical and direct energy conversion, high-energy propellants, electric propulsion and high performance rocket fuels and nozzles have led to significant technical and management roles in many natural space nuclear power and propulsion programs.

Historical perspectives: The role of the NASA Lewis Research Center in the national space nuclear power programs

NASA Technical Reports Server (NTRS)

Bloomfield, H. S.; Sovie, R. J.

1991-01-01

The history of the NASA Lewis Research Center's role in space nuclear power programs is reviewed. Lewis has provided leadership in research, development, and the advancement of space power and propulsion systems. Lewis' pioneering efforts in nuclear reactor technology, shielding, high temperature materials, fluid dynamics, heat transfer, mechanical and direct energy conversion, high-energy propellants, electric propulsion and high performance rocket fuels and nozzles have led to significant technical and management roles in many national space nuclear power and propulsion programs.

The RERTR Program : a status report.

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

Travelli, A.

1998-10-19

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

REACTOR PHYSICS QUARTERLY REPORT JANUARY, FEBRUARY, MARCH 1970

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

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

1970-05-01

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

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

A Roadmap of Innovative Nuclear Energy System

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2017-01-01

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

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

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

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

2011-01-01

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

Light Water Reactor Sustainability Accomplishments Report

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

McCarthy, Kathryn A.

Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80%more » below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you have any questions about the information in the report, or about the LWRS Program, please contact me, Richard A. Reister (the Federal Program Manager), or the respective research pathway leader (noted on pages 26 and 27), or visit the LWRS Program website (www.inl.gov/lwrs). The annually updated Integrated Program Plan and Pathway Technical Program Plans are also available for those seeking more detailed technical Information.« less

Reducing The Nuclear Danger

DTIC Science & Technology

1995-10-01

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

Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance

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

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

2010-01-01

Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transportmore » and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and operational issues and data related to assembly burnup confirmation. The objective of this paper is to summarize the work and significant accomplishments, with references to the technical reports and publications for complete details, and provide a useful resource to others in the burnup credit community.« less

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

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

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

RADSOURCE. Volume 1, Part 1, A scaling factor prediction computer program technical manual and code validation: Final report

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

Vance, J.N.; Holderness, J.H.; James, D.W.

1992-12-01

Waste stream scaling factors based on sampling programs are vulnerable to one or more of the following factors: sample representativeness, analytic accuracy, and measurement sensitivity. As an alternative to sample analyses or as a verification of the sampling results, this project proposes the use of the RADSOURCE code, which accounts for the release of fuel-source radionuclides. Once the release rates of these nuclides from fuel are known, the code develops scaling factors for waste streams based on easily measured Cobalt-60 (Co-60) and Cesium-137 (Cs-137). The project team developed mathematical models to account for the appearance rate of 10CFR61 radionuclides inmore » reactor coolant. They based these models on the chemistry and nuclear physics of the radionuclides involved. Next, they incorporated the models into a computer code that calculates plant waste stream scaling factors based on reactor coolant gamma- isotopic data. Finally, the team performed special sampling at 17 reactors to validate the models in the RADSOURCE code.« less

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

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

E.T. Robinson; John Sirman; Prasad Apte

2005-05-01

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

Program report for FY 1984 and 1985 Atmospheric and Geophysical Sciences Division of the Physics Department

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

Knox, J.B.; MacCracken, M.C.; Dickerson, M.H.

1986-08-01

This annual report for the Atmospheric and Geophysical Sciences Division (G-Division) summarizes the activities and highlights of the past three years, with emphasis on significant research findings in two major program areas: the Atmospheric Release Advisory Capability (ARAC), with its recent involvement in assessing the effects of the Chernobyl reactor accident, and new findings on the environmental consequences of nuclear war. The technical highlights of the many other research projects are also briefly reported, along with the Division's organization, budget, and publications.

Nuclear Reactors and Technology

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

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

1992-01-01

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

Next Generation Nuclear Plant Methods Research and Development Technical Program Plan -- PLN-2498

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

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2008-09-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

Next Generation Nuclear Plant Methods Technical Program Plan

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

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2010-12-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

Next Generation Nuclear Plant Methods Technical Program Plan -- PLN-2498

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

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2010-09-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

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

Hecker, Siegfried S.

Actions of the Government of the Democratic People's Republic of Korea have precipitated two nuclear crises in the past 10 years. The 1994 crisis was resolved through the 'Agreed Framework.' North Korea agreed to 'freeze' and eventually dismantle its nuclear program (with U.S. help to store spent fuel safely and under IAEA inspection). In return, the United States agreed (with the KEDO international consortium) to build two light-water reactors and supply North Korea with heavy-fuel oil until the reactors come on line. In addition, both sides agreed to move towards full normalization of relations, work for peace and security onmore » a nuclear-free Korean Peninsula, and work on strengthening the international nonproliferation regime. The second nuclear crisis erupted when North Korean Government officials allegedly admitted to having a clandestine uranium enrichment program when confronted with this accusation by U.S. officials in October 2002. The United States (through KEDO) suspended heavy-fuel oil shipments and North Korea responded by expelling the IAEA inspectors, withdrawing from the Nuclear Nonproliferation Treaty, and restarting its nuclear program in January 2003. The North Korean Government has invited Professor John Lewis of Stanford University, a China and North Korea scholar, for Track I1 discussions of nuclear and other key issues since 1987. In August 2003, Professor Lewis visited North Korea just before the first six-party talks, which were designed by the United States to solve the current nuclear crisis. Professor Lewis was invited back for the January 2004 visit. He asked Jack Pritchard, former U.S. special envoy for DRPK negotiations, and me to accompany him. Two Asian affairs staff specialists from the U.S. Senate Foreign Relations Committee also joined us. I will report on the visit to the Yongbyon Nuclear Scientific Research Center on January 8,2004. We toured the 5 MWe reactor, the 50 MWe reactor construction site, the spent fuel pool storage building, and the radiochemical laboratory. We concluded that North Korea has restarted its 5 MWe reactor (which produces roughly 6 kg of plutonium annually), it removed the 8000 spent fuel rods that were previously stored under IAEA safeguards from the spent fuel pool, and that it most likely extracted the 25 to 30 kg of plutonium contained in these fuel rods. Although North Korean officials showed us what they claimed was their plutonium metal product from this reprocessing campaign, we were not able to conclude definitively that it was in fact plutonium metal and that it came from the most recent reprocessing campaign. Nevertheless, our North Korean hosts demonstrated that they had the capability, the facility and requisite capacity, and the technical expertise to produce plutonium metal. We were not shown any facilities or had the opportunity to talk to technical or military experts who were able to address the issue of whether or not North Korea had a 'deterrent' as claimed - that is, we were not able to conclude that North Korea can build a nuclear device and that it can integrate nuclear devices into suitable delivery systems. On the matter of uranium enrichment programs, Vice Minister Kim Gye Gwan categorically denied that North Korea has a uranium enrichment program - he said, 'we have no program, no equipment, and no technical expertise for uranium enrichment.' Upon return to the United States, I shared my observations and analysis with U.S. Government officials in Washington, DC, including congressional testimony to the Senate Foreign Relations Committee and briefings to two House of Representative Committees.« less

Light Water Reactor Sustainability Program: Risk-Informed Safety Margins Characterization (RISMC) Pathway Technical Program Plan

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

Smith, Curtis; Rabiti, Cristian; Martineau, Richard

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degreemore » of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”« less

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

DTIC Science & Technology

1986-05-01

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

Nuclear Physics Made Very, Very Easy

NASA Technical Reports Server (NTRS)

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

1968-01-01

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

Five Lectures on Nuclear Reactors Presented at Cal Tech

DOE R&D Accomplishments Database

Weinberg, Alvin M.

1956-02-10

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

Safety system augmentation at Russian nuclear power plants

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

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

1996-12-31

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

Analysis of UF6 breeder reactor power plants

NASA Technical Reports Server (NTRS)

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

1976-01-01

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

77 FR 38742 - Non-Power Reactor License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-29

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

Chemical vapor deposition growth

NASA Technical Reports Server (NTRS)

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

1978-01-01

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

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2013 CFR

2013-01-01

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

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2014 CFR

2014-01-01

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

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2012 CFR

2012-01-01

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

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

Fusion Materials Semiannual Progress Report for Period Ending December 31, 1998

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

Rowcliff, A.F.; Burn, G.

1999-04-01

This is the twenty-fifth in a series of semiannual technical progress reports on fusion materials. This report combines the full spectrum of research and development activities on both metallic and non-metallic materials with primary emphasis on the effects of the neutronic and chemical environment on the properties and performance of materials for in-vessel components. This effort forms one element of the materials program being conducted in support of the Fusion Energy Sciences Program of the U.S. Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reportedmore » separately.« less

Strengthening IAEA Safeguards for Research Reactors

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

Reid, Bruce D.; Anzelon, George A.; Budlong-Sylvester, Kory

During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half amore » dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To broaden the IAEA safeguards toolbox, the study recommends that the Agency consider closing potential gaps in safeguards coverage by, among other things: 1) adapting its safeguards measures based on a case-by-case assessment; 2) using more frequent and expanded/enhanced mailbox declarations (ideally with remote transmission of the data to IAEA Headquarters in Vienna) coupled with short-notice or unannounced inspections; 3) putting more emphasis on the collection and analysis of environmental samples at hot cells and waste storage tanks; 4) taking Safeguards by Design into account for the construction of new research reactors and best practices for existing research reactors; 5) utilizing fully all legal authorities to enhance inspection access (including a strengthened and continuing DIV process); and 6) utilizing new approaches to improve auditing activities, verify reactor operating data history, and track/monitor the movement and storage of spent fuel.« less

The Euratom Seventh Framework Programme FP7 (2007-2011)

NASA Astrophysics Data System (ADS)

Garbil, R.

2010-10-01

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

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.

Implementation Plan for Qualification of Sodium-Cooled Fast Reactor Technology Information

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

Moe, Wayne; Honma, George

This document identifies and discusses implementation elements that can be used to facilitate consistent and systematic evaluation processes relating to quality attributes of technical information (with focus on SFR technology) that will be used to support licensing of advanced reactor designs. 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 approach for determining acceptability of test data, analysis, and/or other technical informationmore » is based on guidance provided in INL/EXT-15-35805, “Guidance on Evaluating Historic Technology Information for Use in Advanced Reactor Licensing.” The implementation plan can be adopted into a working procedure at each of the national laboratories performing data qualification, or by applicants seeking future license application for advanced reactor technology.« less

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

PubMed

Yang, Yiming; Li, Jian; He, Hong

2017-08-24

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

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

Frybort, Jan

A critical experiment is a standard part of training of students at the Training Reactor VR-1 operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague. In autumn 2005 the HEU fuel IRT-3M with enrichment 36 % {sup 235}U was replaced by the LEU fuel IRT-4M with enrichment 19.7 % {sup 235}U. The fuel replacement at the VR-1 Reactor is a part of an international program RERTR. This Paper presents basic information about preparation for the fuel replacement and approaching of the first critical state with the new zone configuration C1 which replacedmore » B1 core with the old IRT-3M fuel. The whole process was carried out according to the Czech law and the relevant international recommendations. The experience with the VR-1 operation confirms the assumption that the C1 core configuration will be suitable from the point of view of the reactivity balance for the long term safe operation of the Training Reactor VR-1. (author)« less

In-vessel melt retention as a severe accident management strategy for the Loviisa Nuclear Power Plant

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

Kymaelaeinen, O.; Tuomisto, H.; Theofanous, T.G.

1997-02-01

The concept of lower head coolability and in-vessel retention of corium has been approved as a basic element of the severe accident management strategy for IVO`s Loviisa Plant (VVER-440) in Finland. The selected approach takes advantage of the unique features of the plant such as low power density, reactor pressure vessel without penetrations at the bottom and ice-condenser containment which ensures flooded cavity in all risk significant sequences. The thermal analyses, which are supported by experimental program, demonstrate that in Loviisa the molten corium on the lower head of the reactor vessel is coolable externally with wide margins. This papermore » summarizes the approach and the plant modifications being implemented. During the approval process some technical concerns were raised, particularly with regard to thermal loadings caused by contact of cool cavity water and hot corium with the reactor vessel. Resolution of these concerns is also discussed.« less

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

Not Available

Gulf Research and Development Company is implementing a DOE-sponsored Underground Coal Gasification project in Steeply Dipping Coal Beds (UCG/SDB) in order to assess the economic and technical viability of UCG in SDB. In the Fall 1980 drilling program, 2 vertical and 2 slant process wells; 3 hydrologic and 1 exploratory well and 4 HFEM wells were completed. The Spring, 1981 program will consist of drilling the remaining instrumentation wells necessary to track the progress of the underground reactor in real time. These will consist of: 6 additional High Frequency Electromagnetic wells (HFEM) and 3 extensometer wells (X). These wells willmore » be installed vertically with an expected deviation of two degrees or less.« less

Richland five-year O2 R and D Program. Integrated site operation

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

Not Available

1966-07-11

The technical feasibility of using an electrolytic reduction process to reduce metal scrap and oxide to usable uranium metal is being studied. The incentives for using electrolytic reduction at Richland may be summarized as follows: (1) reduce the unit and total costs of producing plutonium; (2) increase the flexibility of the Richland reactors for producing isotopes, particularly U-236; and (3) simplify the present fuel cycle complex. The scope of the mission is limited to the evaluation of hollow extruded I and E cores, the evaluation of electro-reduced uranium, an investigation of the solution rate of UO{sub 2} in the electrolyte,more » and small-scale irradiations of UO{sub 2} fuels in the N and K Reactors. Progress during FY 1966 is summarized.« less

Reference reactor module for NASA's lunar surface fission power system

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

Poston, David I; Kapernick, Richard J; Dixon, David D

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on themore » lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.« less

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.

10 CFR 54.21 - Contents of application-technical information.

Code of Federal Regulations, 2011 CFR

2011-01-01

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

10 CFR 54.21 - Contents of application-technical information.

Code of Federal Regulations, 2010 CFR

2010-01-01

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

TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

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

Ryan, B.C.

1997-05-01

This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely themore » total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.« less

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

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

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

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

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

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

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

Corradini, M. L.

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

Twenty-fifth water reactor safety information meeting: Proceedings. Volume 2: Human reliability analysis and human performance evaluation; Technical issues related to rulemakings; Risk-informed, performance-based initiatives; High burn-up fuel research

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

Monteleone, S.

1998-03-01

This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) human reliability analysis and human performance evaluation; (2) technical issues relatedmore » to rulemakings; (3) risk-informed, performance-based initiatives; and (4) high burn-up fuel research. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.« less

Some methods for achieving more efficient performance of fuel assemblies

NASA Astrophysics Data System (ADS)

Boltenko, E. A.

2014-07-01

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

10 CFR 72.24 - Contents of application: Technical information.

Code of Federal Regulations, 2011 CFR

2011-01-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-27

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

Selection of specimen types for irradiation surveillance programs

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

Varga, T.; Njo, D.H.

1981-10-01

Recent trends in coping with embrittlement problems in reactor pressure vessels (RPVs) show two main directions of development: (1) improvement of the vessel materials and (2) limitations of fluence over the design life of the RPV. For several reasons, however, adequate irradiation surveillance programs are still considered to be necessary in the future, despite possible improvements resulting from such research activities. Since the introduction of linear elastic fracture mechanics (LEFM) and elastic-plastic fracture mechanics, (EPFM), irradiation surveillance programs show a trend towards direct measurement of fracture toughness, in addition to relying on the conventional nil-ductility transition temperature (NDTT) shift asmore » a relative measure of embrittlement. Some basic considerations concerning the selection of specimen types for irradiation surveillance programs and some technical aspects of currently used speciment types are discussed.« less

Text processing for technical reports (direct computer-assisted origination, editing, and output of text)

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

De Volpi, A.; Fenrick, M. R.; Stanford, G. S.

1980-10-01

Documentation often is a primary residual of research and development. Because of this important role and because of the large amount of time consumed in generating technical reports, particularly those containing formulas and graphics, an existing data-processing computer system has been adapted so as to provide text-processing of technical documents. Emphasis has been on accuracy, turnaround time, and time savings for staff and secretaries, for the types of reports normally produced in the reactor development program. The computer-assisted text-processing system, called TXT, has been implemented to benefit primarily the originator of technical reports. The system is of particular value tomore » professional staff, such as scientists and engineers, who have responsibility for generating much correspondence or lengthy, complex reports or manuscripts - especially if prompt turnaround and high accuracy are required. It can produce text that contains special Greek or mathematical symbols. Written in FORTRAN and MACRO, the program TXT operates on a PDP-11 minicomputer under the RSX-11M multitask multiuser monitor. Peripheral hardware includes videoterminals, electrostatic printers, and magnetic disks. Either data- or word-processing tasks may be performed at the terminals. The repertoire of operations has been restricted so as to minimize user training and memory burden. Spectarial staff may be readily trained to make corrections from annotated copy. Some examples of camera-ready copy are provided.« less

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-08

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

Nuclear powerplants for mobile applications.

NASA Technical Reports Server (NTRS)

Anderson, J. L.

1972-01-01

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

Nuclear power plants for mobile applications

NASA Technical Reports Server (NTRS)

Anderson, J. L.

1972-01-01

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

Aluminum Carbothermic Technology

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

Bruno, Marshall J.

2005-03-31

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-04

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

Development plan for the External Hazards Experimental Group. Light Water Reactor Sustainability Program

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

Coleman, Justin Leigh; Smith, Curtis Lee; Burns, Douglas Edward

This report describes the development plan for a new multi-partner External Hazards Experimental Group (EHEG) coordinated by Idaho National Laboratory (INL) within the Risk-Informed Safety Margin Characterization (RISMC) technical pathway of the Light Water Reactor Sustainability Program. Currently, there is limited data available for development and validation of the tools and methods being developed in the RISMC Toolkit. The EHEG is being developed to obtain high-quality, small- and large-scale experimental data validation of RISMC tools and methods in a timely and cost-effective way. The group of universities and national laboratories that will eventually form the EHEG (which is ultimately expectedmore » to include both the initial participants and other universities and national laboratories that have been identified) have the expertise and experimental capabilities needed to both obtain and compile existing data archives and perform additional seismic and flooding experiments. The data developed by EHEG will be stored in databases for use within RISMC. These databases will be used to validate the advanced external hazard tools and methods.« less

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

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

Pan, Paul Y

2010-12-10

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

Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

NASA Astrophysics Data System (ADS)

Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

2017-07-01

Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

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

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

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

2015-05-01

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

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

NASA Astrophysics Data System (ADS)

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

2008-10-01

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

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

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

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

Structural Materials and Fuels for Space Power Plants

NASA Technical Reports Server (NTRS)

Bowman, Cheryl; Busby, Jeremy; Porter, Douglas

2008-01-01

A fission reactor combined with Stirling convertor power generation is one promising candidate in on-going Fission Surface Power (FSP) studies for future lunar and Martian bases. There are many challenges for designing and qualifying space-rated nuclear power plants. In order to have an affordable and sustainable program, NASA and DOE designers want to build upon the extensive foundation in nuclear fuels and structural materials. This talk will outline the current Fission Surface Power program and outline baseline design options for a lunar power plant with an emphasis on materials challenges. NASA first organized an Affordable Fission Surface Power System Study Team to establish a reference design that could be scrutinized for technical and fiscal feasibility. Previous papers and presentations have discussed this study process in detail. Considerations for the reference design included that no significant nuclear technology, fuels, or material development were required for near term use. The desire was to build upon terrestrial-derived reactor technology including conventional fuels and materials. Here we will present an overview of the reference design, Figure 1, and examine the materials choices. The system definition included analysis and recommendations for power level and life, plant configuration, shielding approach, reactor type, and power conversion type. It is important to note that this is just one concept undergoing refinement. The design team, however, understands that materials selection and improvement must be an integral part of the system development.

Joint Assessment of Renewable Energy and Water Desalination Research Center (REWDC) Program Capabilities and Facilities In Radioactive Waste Management

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

Bissani, M; Fischer, R; Kidd, S

2006-04-03

The primary goal of this visit was to perform a joint assessment of the Renewable Energy and Water Desalination Center's (REWDC) program in radioactive waste management. The visit represented the fourth technical and scientific interaction with Libya under the DOE/NNSA Sister Laboratory Arrangement. Specific topics addressed during the visit focused on Action Sheet P-05-5, ''Radioactive Waste Management''. The Team, comprised of Mo Bissani (Team Lead), Robert Fischer, Scott Kidd, and Jim Merrigan, consulted with REWDC management and staff. The team collected information, discussed particulars of the technical collaboration and toured the Tajura facility. The tour included the waste treatment facility,more » waste storage/disposal facility, research reactor facility, hot cells and analytical labs. The assessment team conducted the first phase of Task A for Action Sheet 5, which involved a joint assessment of the Radioactive Waste Management Program. The assessment included review of the facilities dedicated to the management of radioactive waste at the Tourja site, the waste management practices, proposed projects for the facility and potential impacts on waste generation and management.« less

Acceptability of reactors in space

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

Buden, D.

1981-04-01

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

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.

Current biotechnological developments in Belgium.

PubMed

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

1989-01-01

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

Approach to developing reliable space reactor power systems

NASA Technical Reports Server (NTRS)

Mondt, Jack F.; Shinbrot, Charles H.

1991-01-01

During Phase II, the Engineering Development Phase, the SP-100 Project has defined and is pursuing a new approach to developing reliable power systems. The approach to developing such a system during the early technology phase is described along with some preliminary examples to help explain the approach. Developing reliable components to meet space reactor power system requirements is based on a top-down systems approach which includes a point design based on a detailed technical specification of a 100-kW power system. The SP-100 system requirements implicitly recognize the challenge of achieving a high system reliability for a ten-year lifetime, while at the same time using technologies that require very significant development efforts. A low-cost method for assessing reliability, based on an understanding of fundamental failure mechanisms and design margins for specific failure mechanisms, is being developed as part of the SP-100 Program.

Superconducting magnet development for tokamaks and mirrors: a technical assessment

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

Laverick, C.; Jacobs, R. B.; Boom, R. W.

1977-11-01

The role of superconducting magnets in Magnetic Fusion Energy Research and Development is assessed from a consideration of program plans and schedules, the present status of the programs and the research and development suggestions arising from recent studies and workshops. A principal conclusion is that the large superconducting magnet systems needed for commercial magnetic fusion reactors can be constructed. However such magnets working under severe conditions, with increasingly stringent reliability, safety and cost restrictions can never be built unless experience is first gained in a number of important installations designed to prove physics and technology steps on the way tomore » commercial power demonstration. The immediate problem is to design a technology program in the absence of definite device needs and specifications, giving a priority weighting to the multiplicity of good, high quality development program suggestions when all proposals cannot be supported.« less

Imaging Fukushima Daiichi reactors with muons

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

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.

2013-05-15

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi tomore » make this determination in the near future.« less

Imaging Fukushima Daiichi reactors with muons

NASA Astrophysics Data System (ADS)

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Lukić, Zarija; Masuda, Koji; Milner, Edward C.; Morris, Christopher L.; Perry, John O.

2013-05-01

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

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

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

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

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

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

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

1991-09-01

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

Office for Analysis and Evaluation of Operational Data 1996 annual report. Volume 10, Number 1: Reactors

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

NONE

1997-12-01

This annual report of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) describes activities conducted during 1996. The report is published in three parts. NUREG-1272, Vol. 10, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year and summarizes information from such sources as licensee event reports and reports to the NRC`s Operations Center. NUREG-1272, Vol.more » 10, No. 2, covers nuclear materials and presents a review of the events and concerns during 1996 associated with the use of licensed material in nonreactor applications, such as personnel overexposures and medical misadministrations. Both reports also contain a discussion of the Incident Investigation Team program and summarize both the Incident Investigation Team and Augmented Inspection Team reports. Each volume contains a list of the AEOD reports issued from CY 1980 through 1996. NUREG-1272, Vol. 10, No. 3, covers technical training and presents the activities of the Technical Training Center in support of the NRC`s mission in 1996.« less

Chemical Technology Division, Annual technical report, 1991

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

Not Available

1992-03-01

Highlights of the Chemical Technology (CMT) Division's activities during 1991 are presented. In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for fluidized-bed combustion and coal-fired magnetohydrodynamics; (3) methods for treatment of hazardous and mixed hazardous/radioactive waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for an unsaturated repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (IFR); (7) processes for removalmore » of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials in environments simulating those of fission and fusion energy systems. The Division also conducts basic research in catalytic chemistry associated with molecular energy resources; chemistry of superconducting oxides and other materials of interest with technological application; interfacial processes of importance to corrosion science, catalysis, and high-temperature superconductivity; and the geochemical processes involved in water-rock interactions occurring in active hydrothermal systems. In addition, the Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the technical programs at Argonne National Laboratory (ANL).« less

Chemical Technology Division, Annual technical report, 1991

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

Not Available

1992-03-01

Highlights of the Chemical Technology (CMT) Division`s activities during 1991 are presented. In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for fluidized-bed combustion and coal-fired magnetohydrodynamics; (3) methods for treatment of hazardous and mixed hazardous/radioactive waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for an unsaturated repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (IFR); (7) processes for removalmore » of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials in environments simulating those of fission and fusion energy systems. The Division also conducts basic research in catalytic chemistry associated with molecular energy resources; chemistry of superconducting oxides and other materials of interest with technological application; interfacial processes of importance to corrosion science, catalysis, and high-temperature superconductivity; and the geochemical processes involved in water-rock interactions occurring in active hydrothermal systems. In addition, the Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the technical programs at Argonne National Laboratory (ANL).« less

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

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

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

2009-10-09

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

Reference Reactor Module for the Affordable Fission Surface Power System

NASA Astrophysics Data System (ADS)

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

2008-01-01

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

Saturn Apollo Program

NASA Image and Video Library

1967-09-09

This is the official NASA portrait of astronaut William Anders. Anders was commissioned in the air Force after graduation from the Naval Academy and served as a fighter pilot in all-weather interception squadrons of the Air Defense Command. Later he was responsible for technical management of nuclear power reactor shielding and radiation effects programs while at the Air Force Weapons Laboratory in New Mexico. In 1964, Anders was selected by the National Aeronautics and Space Administration (NASA) as an astronaut with responsibilities for dosimetry, radiation effects and environmental controls. He was backup pilot for the Gemini XI, Apollo 11 flights, and served as lunar module (LM) pilot for Apollo 8, the first lunar orbit mission in December 1968. He has logged more than 6,000 hours flying time.

Space fusion energy conversion using a field reversed configuration reactor: A new technical approach for space propulsion and power

NASA Technical Reports Server (NTRS)

Schulze, Norman R.; Miley, George H.; Santarius, John F.

1991-01-01

The fusion energy conversion design approach, the Field Reversed Configuration (FRC) - when burning deuterium and helium-3, offers a new method and concept for space transportation with high energy demanding programs, like the Manned Mars Mission and planetary science outpost missions require. FRC's will increase safety, reduce costs, and enable new missions by providing a high specific power propulsion system from a high performance fusion engine system that can be optimally designed. By using spacecraft powered by FRC's the space program can fulfill High Energy Space Missions (HESM) in a manner not otherwise possible. FRC's can potentially enable the attainment of high payload mass fractions while doing so within shorter flight times.

Innovations for In-Pile Measurements in the Framework of the CEA-SCK•CEN Joint Instrumentation Laboratory

NASA Astrophysics Data System (ADS)

Villard, Jean-Francois; Schyns, Marc

2010-12-01

Optimizing the life cycle of nuclear systems under safety constraints requires high-performance experimental programs to reduce uncertainties on margins and limits. In addition to improvement in modeling and simulation, innovation in instrumentation is crucial for analytical and integral experiments conducted in research reactors. The quality of nuclear research programs relies obviously on an excellent knowledge of their experimental environment which constantly calls for better online determination of neutron and gamma flux. But the combination of continuously increasing scientific requirements and new experimental domains -brought for example by Generation IV programsnecessitates also major innovations for in-pile measurements of temperature, dimensions, pressure or chemical analysis in innovative mediums. At the same time, the recent arising of a European platform around the building of the Jules Horowitz Reactor offers new opportunities for research institutes and organizations to pool their resources in order to face these technical challenges. In this situation, CEA (French Nuclear Energy Commission) and SCK'CEN (Belgian Nuclear Research Centre) have combined their efforts and now share common developments through a Joint Instrumentation Laboratory. Significant progresses have thus been obtained recently in the field of in-pile measurements, on one hand by improvement of existing measurement methods, and on the other hand by introduction in research reactors of original measurement techniques. This paper highlights the state-of-the-art and the main requirements regarding in-pile measurements, particularly for the needs of current and future irradiation programs performed in material testing reactors. Some of the main on-going developments performed in the framework of the Joint Instrumentation Laboratory are also described, such as: - a unique fast neutron flux measurement system using fission chambers with 242Pu deposit and a specific online data processing, - an optical system designed to perform in-pile dimensional measurements of material samples under irradiation, - an acoustical instrumentation allowing the online characterization of fission gas release in Pressurized Water Reactor fuel rods. For each example, the obtained results, expected impacts and development status are detailed.

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

NASA Technical Reports Server (NTRS)

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

2003-01-01

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

SAM Theory Manual

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

Hu, Rui

The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM development aims for advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability for reactor transient analyses. To facilitate the code development, SAM utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage modern advanced software environments and numerical methods. SAM focuses on modeling advanced reactormore » concepts such as SFRs (sodium fast reactors), LFRs (lead-cooled fast reactors), and FHRs (fluoride-salt-cooled high temperature reactors) or MSRs (molten salt reactors). These advanced concepts are distinguished from light-water reactors in their use of single-phase, low-pressure, high-temperature, and low Prandtl number (sodium and lead) coolants. As a new code development, the initial effort has been focused on modeling and simulation capabilities of heat transfer and single-phase fluid dynamics responses in Sodium-cooled Fast Reactor (SFR) systems. The system-level simulation capabilities of fluid flow and heat transfer in general engineering systems and typical SFRs have been verified and validated. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As the code is still under ongoing development, this SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.« less

Space nuclear reactors — A post-operational disposal strategy

NASA Astrophysics Data System (ADS)

Angelo, Joseph A.; Buden, David

If 100-kWe and multimegawatt-electric class space nuclear reactors are to play a significant role in humanity's push into cislunar and heliocentric space in the next millennium, the obvious advantages of space nuclear power plants should not be denied to space mission planners due to a failure to develop internationally-acceptable post-operational disposal strategies for spent reactor cores. This is true whether the space reactor has shut down at the end of its normal mission lifetime or in response to an onboard system failure/emergency which causes a premature mission termination. Up until now the great majority of aerospace nuclear safety efforts have concentrated on prelaunch, launch and reactor startup activities. In fact, with the exception of the development of the "nuclear safe orbit" (NSO) concept, little technical attention has yet been given to the post-operational disposal of future space reactors. This paper describes the technical alternatives available for the safe, acceptable disposal of space reactors that could be used in a wide variety of space applications in the 21st Century. Post-operational core radioactivity levels for typical advanced design (hundred kWe-class) space reactors are presented as a function of decay time and contrasted to the spent core radionuclide inventory of the SNAP-10A system, the only nuclear reactor operated in space by the United States. The role of a permanent space station, smart robotic systems, and an operating lunar base in support of spent core disposal strategies is also presented, including use of a selected portion of the lunar surface as an internationally-designated spent reactor core repository.

Conceptual design and issues of the laser inertial fusion test (LIFT) reactor—targets and chamber systems

NASA Astrophysics Data System (ADS)

Norimatsu, T.; Kozaki, Y.; Shiraga, H.; Fujita, H.; Okano, K.; Members of LIFT Design Team

2017-11-01

We present the conceptual design of an experimental laser fusion plant known as the laser inertial fusion test (LIFT) reactor. The conceptual design aims at technically connecting a single-shot experiment and a commercial power plant. The LIFT reactor is designed on a three-phase scheme, where each phase has specific goals and the dedicated chambers of each phase are driven by the same laser. Technical issues related to the chamber technology including radiation safety to repeat burst mode operation are discussed in this paper.

Technical Readiness and Gaps Analysis of Commercial Optical Materials and Measurement Systems for Advanced Small Modular Reactors

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

Anheier, Norman C.; Suter, Jonathan D.; Qiao, Hong

2013-08-06

This report intends to support Department of Energy’s Office of Nuclear Energy (DOE-NE) Nuclear Energy Research and Development Roadmap and industry stakeholders by evaluating optical-based instrumentation and control (I&C) concepts for advanced small modular reactor (AdvSMR) applications. These advanced designs will require innovative thinking in terms of engineering approaches, materials integration, and I&C concepts to realize their eventual viability and deployability. The primary goals of this report include: 1. Establish preliminary I&C needs, performance requirements, and possible gaps for AdvSMR designs based on best available published design data. 2. Document commercial off-the-shelf (COTS) optical sensors, components, and materials in termsmore » of their technical readiness to support essential AdvSMR in-vessel I&C systems. 3. Identify technology gaps by comparing the in-vessel monitoring requirements and environmental constraints to COTS optical sensor and materials performance specifications. 4. Outline a future research, development, and demonstration (RD&D) program plan that addresses these gaps and develops optical-based I&C systems that enhance the viability of future AdvSMR designs. The development of clean, affordable, safe, and proliferation-resistant nuclear power is a key goal that is documented in the Nuclear Energy Research and Development Roadmap. This roadmap outlines RD&D activities intended to overcome technical, economic, and other barriers, which currently limit advances in nuclear energy. These activities will ensure that nuclear energy remains a viable component to this nation’s energy security.« less

Self-teaching neural network learns difficult reactor control problem

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

Jouse, W.C.

1989-01-01

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

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

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

Betzler, Benjamin R; Mays, Gary T

2016-01-01

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

Helium refrigerator maintenance and reliability at the OPAL cold neutron source

NASA Astrophysics Data System (ADS)

Thiering, Russell; Taylor, David; Lu, Weijian

2012-06-01

Australia's first Cold Neutron Source (CNS) is a major asset to its nuclear research program. The CNS, and associated helium refrigerator, was commissioned in 2006 and is operated at the Open Pool Light Water nuclear Reactor (OPAL). The OPAL CNS operates a 20K, 5 kW Brayton cycle helium refrigerator. In this paper relevant experiences from helium refrigerator operation, maintenance and repair are presented along with the lessons learnt from a series of technical investigations. Turbine failure, due to volatile organic species, is discussed along with the related compressor oil degradation and oil separation efficiency.

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

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

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

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

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

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

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

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

Karpov, A. S.

2013-01-15

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-22

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

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.

Preparation of the Second Shipment of Spent Nuclear Fuel from the Ustav Jaderneho Vyzkumu Rez (UJV Rez), a.s., Czech Republic to the Russian Federation for Reprocessing - 13478

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

Trtilek, Radek; Podlaha, Josef

After more than 50 years of operation of the LVR-15 research reactor operated by the UJV Rez, a. s. (formerly Nuclear Research Institute - NRI), a large amount of the spent nuclear fuel (SNF) of Russian origin has been accumulated. In 2005 UJV Rez, a. s. jointed the Russian Research Reactor Fuel Return (RRRFR) program under the United States (US) - Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). In 2007 the first shipment of SNF was realized. In 2011, preparation of the secondmore » shipment of spent fuel from the Czech Republic started. The experience obtained from the first shipment will be widely used, but some differences must be taken into the account. The second shipment will be realized in 2013 and will conclude the return transport of all, both fresh and spent, high-enriched nuclear fuel from the Czech Republic to the Russian Federation. After the shipment is completed, there will be only low-enriched nuclear fuel on the territory of the Czech Republic, containing maximum of 20% of U-235, which is the conventionally recognized limit between the low- and high-enriched nuclear materials. The experience (technical, organizational, administrative, logistic) obtained from the each SNF shipment as from the Czech Republic as from other countries using the Russian type research reactors are evaluated and projected onto preparation of next shipment of high enriched nuclear fuel back to the Russian Federation. The results shown all shipments provided by the UJV Rez, a. s. in the frame of the GTRI Program have been performed successfully and safely. It is expected the experience and results will be applied to preparation and completing of the Chinese Miniature Neutron Source Reactors (MNSR) Spent Nuclear Fuel Repatriation in the near future. (authors)« less

International-Aerial Measuring System (I-AMS) Training Program

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

Wasiolek, Piotre T.; Malchor, Russell L.; Maurer, Richard J.

2015-10-01

Since the Fukushima reactor accident in 2011, there has been an increased interest worldwide in developing national capabilities to rapidly map and assess ground contamination resulting from nuclear reactor accidents. The capability to rapidly measure the size of the contaminated area, determine the activity level, and identify the radionuclides can aid emergency managers and decision makers in providing timely protective action recommendations to the public and first responders. The development of an aerial detection capability requires interagency coordination to assemble the radiation experts, detection system operators, and aviation aircrews to conduct the aerial measurements, analyze and interpret the data, andmore » provide technical assessments. The Office of International Emergency Management and Cooperation (IEMC) at the U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) sponsors an International - Aerial Measuring System (I-AMS) training program for partner nations to develop and enhance their response to radiological emergencies. An initial series of courses can be conducted in the host country to assist in developing an aerial detection capability. As the capability develops and expands, additional experience can be gained through advanced courses with the opportunity to conduct aerial missions over a broad range of radiation environments.« less

Progress on RERTR activities in Argentina

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

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

2008-07-15

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

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

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

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

Bissani, M; Tyson, S

Vietnam's nuclear program began in the 1960s with the installation at Dalat of a 250 kW TRIGA Mk-II research reactor under the U.S. Atoms for Peace Program. The reactor was shut down and its core removed only a few years later, and the nuclear research program was suspended until after the end of the civil war in the late 1970s. The Soviet Union assisted Vietnam in restoring the Dalat reactor to an operational status in 1984, trained a cadre of scientific and technical staff in its operation, and contributed to the development of nuclear science for the medical and agriculturalmore » sectors. In the agricultural area in particular, Vietnamese experts have been very successful in developing mutant strains of rice, and continue to work with the IAEA to yield strains that have a shorter growing period, increased resistance to disease, and other desirable characteristics. Rice has always been the main crop in Vietnam, but technical cooperation with the IAEA and other states has enabled the country to become one of the top rice producers in the world, exporting much of its annual crop to over two dozen countries annually. More recently, Vietnam's government has shown increasing interest in developing a civil nuclear program to supplement its fossil fuel and other energy resources. Projections from a variety of open sources, ranging from the IAEA, the U.S. Department of Energy's Energy Information Administration (EIA), the Vietnamese government, energy corporations, and think tanks all predict a massive increase in energy consumption--especially electricity--within Vietnam and the region as a whole. This growth in consumption will require a corresponding increase in energy production, which in Vietnam is currently satisfied mainly by fossil fuels (coal) and renewable energy (hydropower and biomass); Vietnam has a refining capacity of about 800 barrels/day. Most of its crude oil is exported to generate export income, and is not used to generate electricity. Although Vietnam is able to meet most of its needs through its own resources, it consumes more electricity than it produces (approximately six billion KWh/a). Open sources indicate that increasing exports of manufactured goods and a corresponding growth in the industrial sector (16% since 2003) will lead to a greater energy shortfall beyond the next decade--between 35 and 65 billion KWh/year after 2020. As a signatory to the Kyoto Treaty and other regional environmental accords, and seeking an alternative to its stretched hydroelectric resources, Vietnam is looking for a means to increase energy production while minimizing emissions of greenhouse gases.« less

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

NASA Astrophysics Data System (ADS)

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

1991-01-01

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

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

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

Jacobsen, George; Back, Christina

2015-10-30

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

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

DTIC Science & Technology

2010-07-09

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

Management of the aging of critical safety-related concrete structures in light-water reactor plants

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

Naus, D.J.; Oland, C.B.; Arndt, E.G.

1990-01-01

The Structural Aging Program has the overall objective of providing the USNRC with an improved basis for evaluating nuclear power plant safety-related structures for continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technology, and quantitative methodology for continued-service determinations. Objectives, accomplishments, and planned activities under each of these tasks are presented. Major program accomplishments include development of a materials property data base for structural materials as well as an aging assessment methodology for concrete structures in nuclear power plants. Furthermore, a review and assessment of inservice inspection techniquesmore » for concrete materials and structures has been complete, and work on development of a methodology which can be used for performing current as well as reliability-based future condition assessment of concrete structures is well under way. 43 refs., 3 tabs.« less

Technical solutions to nonproliferation challenges

NASA Astrophysics Data System (ADS)

Satkowiak, Lawrence

2014-05-01

The threat of nuclear terrorism is real and poses a significant challenge to both U.S. and global security. For terrorists, the challenge is not so much the actual design of an improvised nuclear device (IND) but more the acquisition of the special nuclear material (SNM), either highly enriched uranium (HEU) or plutonium, to make the fission weapon. This paper provides two examples of technical solutions that were developed in support of the nonproliferation objective of reducing the opportunity for acquisition of HEU. The first example reviews technologies used to monitor centrifuge enrichment plants to determine if there is any diversion of uranium materials or misuse of facilities to produce undeclared product. The discussion begins with a brief overview of the basics of uranium processing and enrichment. The role of the International Atomic Energy Agency (IAEA), its safeguard objectives and how the technology evolved to meet those objectives will be described. The second example focuses on technologies developed and deployed to monitor the blend down of 500 metric tons of HEU from Russia's dismantled nuclear weapons to reactor fuel or low enriched uranium (LEU) under the U.S.-Russia HEU Purchase Agreement. This reactor fuel was then purchased by U.S. fuel fabricators and provided about half the fuel for the domestic power reactors. The Department of Energy established the HEU Transparency Program to provide confidence that weapons usable HEU was being blended down and thus removed from any potential theft scenario. Two measurement technologies, an enrichment meter and a flow monitor, were combined into an automated blend down monitoring system (BDMS) and were deployed to four sites in Russia to provide 24/7 monitoring of the blend down. Data was downloaded and analyzed periodically by inspectors to provide the assurances required.

Defining the "proven technology" technical criterion in the reactor technology assessment for Malaysia's nuclear power program

NASA Astrophysics Data System (ADS)

Anuar, Nuraslinda; Kahar, Wan Shakirah Wan Abdul; Manan, Jamal Abdul Nasir Abd

2015-04-01

Developing countries that are considering the deployment of nuclear power plants (NPPs) in the near future need to perform reactor technology assessment (RTA) in order to select the most suitable reactor design. The International Atomic Energy Agency (IAEA) reported in the Common User Considerations (CUC) document that "proven technology" is one of the most important technical criteria for newcomer countries in performing the RTA. The qualitative description of five desired features for "proven technology" is relatively broad and only provides a general guideline to its characterization. This paper proposes a methodology to define the "proven technology" term according to a specific country's requirements using a three-stage evaluation process. The first evaluation stage screens the available technologies in the market against a predefined minimum Technology Readiness Level (TRL) derived as a condition based on national needs and policy objectives. The result is a list of technology options, which are then assessed in the second evaluation stage against quantitative definitions of CUC desired features for proven technology. The potential technology candidates produced from this evaluation is further narrowed down to obtain a list of proven technology candidates by assessing them against selected risk criteria and the established maximum allowable total score using a scoring matrix. The outcome of this methodology is the proven technology candidates selected using an accurate definition of "proven technology" that fulfills the policy objectives, national needs and risk, and country-specific CUC desired features of the country that performs this assessment. A simplified assessment for Malaysia is carried out to demonstrate and suggest the use of the proposed methodology. In this exercise, ABWR, AP1000, APR1400 and EPR designs assumed the top-ranks of proven technology candidates according to Malaysia's definition of "proven technology".

Technical solutions to nonproliferation challenges

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

Satkowiak, Lawrence

2014-05-09

The threat of nuclear terrorism is real and poses a significant challenge to both U.S. and global security. For terrorists, the challenge is not so much the actual design of an improvised nuclear device (IND) but more the acquisition of the special nuclear material (SNM), either highly enriched uranium (HEU) or plutonium, to make the fission weapon. This paper provides two examples of technical solutions that were developed in support of the nonproliferation objective of reducing the opportunity for acquisition of HEU. The first example reviews technologies used to monitor centrifuge enrichment plants to determine if there is any diversionmore » of uranium materials or misuse of facilities to produce undeclared product. The discussion begins with a brief overview of the basics of uranium processing and enrichment. The role of the International Atomic Energy Agency (IAEA), its safeguard objectives and how the technology evolved to meet those objectives will be described. The second example focuses on technologies developed and deployed to monitor the blend down of 500 metric tons of HEU from Russia's dismantled nuclear weapons to reactor fuel or low enriched uranium (LEU) under the U.S.-Russia HEU Purchase Agreement. This reactor fuel was then purchased by U.S. fuel fabricators and provided about half the fuel for the domestic power reactors. The Department of Energy established the HEU Transparency Program to provide confidence that weapons usable HEU was being blended down and thus removed from any potential theft scenario. Two measurement technologies, an enrichment meter and a flow monitor, were combined into an automated blend down monitoring system (BDMS) and were deployed to four sites in Russia to provide 24/7 monitoring of the blend down. Data was downloaded and analyzed periodically by inspectors to provide the assurances required.« less

Guidelines for Implementation of an Advanced Outage Control Center to Improve Outage Coordination, Problem Resolution, and Outage Risk Management

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

St. Germain, Shawn W.; Farris, Ronald K.; Whaley, April M.

This research effort is a part of the Light-Water Reactor Sustainability (LWRS) Program, which is a research and development (R&D) program sponsored by Department of Energy (DOE) and performed in close collaboration with industry R&D programs that provide the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants. The LWRS program serves to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. The purpose of this research is to improve management of nuclear powermore » plant (NPP) outages through the development of an advanced outage control center (AOCC) that is specifically designed to maximize the usefulness of communication and collaboration technologies for outage coordination and problem resolution activities. This technical report for industry implementation outlines methods and considerations for the establishment of an AOCC. This report provides a process for implementation of a change management plan, evaluation of current outage processes, the selection of technology, and guidance for the implementation of the selected technology. Methods are presented for both adoption of technologies within an existing OCC and for a complete OCC replacement, including human factors considerations for OCC design and setup.« less

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

NASA Astrophysics Data System (ADS)

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

2012-06-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-29

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

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

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

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

2011-07-18

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

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

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

Code of Federal Regulations, 2011 CFR

2011-10-01

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

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

Code of Federal Regulations, 2010 CFR

2010-10-01

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

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

Code of Federal Regulations, 2013 CFR

2013-10-01

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

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

Code of Federal Regulations, 2014 CFR

2014-10-01

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

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

Code of Federal Regulations, 2012 CFR

2012-10-01

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

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

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

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

1992-03-01

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

10 CFR 52.137 - Contents of applications; technical information.

Code of Federal Regulations, 2010 CFR

2010-01-01

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

Fusion energy

NASA Astrophysics Data System (ADS)

1990-09-01

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

Chemical Technology Division annual technical report, 1992

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

Battles, J.E.; Myles, K.M.; Laidler, J.J.

1993-06-01

In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for fluidized-bed combustion and coal-fired magnetohydrodynamics; (3) methods for treatment of hazardous waste, mixed hazardous/radioactive waste, and municipal solid waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for an unsaturated repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams, treating water contaminated with volatile organics, and concentrating radioactive waste streams; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (EFR); (7)more » processes for removal of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials (corium; Fe-U-Zr, tritium in LiAlO{sub 2} in environments simulating those of fission and fusion energy systems. The Division also conducts basic research in catalytic chemistry associated with molecular energy resources and novel` ceramic precursors; materials chemistry of superconducting oxides, electrified metal/solution interfaces, and molecular sieve structures; and the geochemical processes involved in water-rock interactions occurring in active hydrothermal systems. In addition, the Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the technical programs at Argonne National Laboratory (ANL).« less

SP-100 Program: space reactor system and subsystem investigations

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

Harty, R.B.

1983-09-30

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

SP-100 program: Space reactor system and subsystem investigations

NASA Astrophysics Data System (ADS)

Harty, R. B.

1983-09-01

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. The nuclear safety review/approval process that is required for a space reactor system is summarized. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that is expected and to provide information that could be usable in future programs.

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

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

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

2000-10-01

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

Current status of U{sub 3}Si{sub 2} fuel element fabrication in Brazil

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

Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.

2008-07-15

IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U{sub 3}O{sub 8}-Al dispersion fuel plates with 2.3 gU/cm{sup 3}. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicidemore » was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)« less

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

Code of Federal Regulations, 2010 CFR

2010-01-01

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

Future prospects for measurements of mass hierarchy and CP violation

NASA Astrophysics Data System (ADS)

Lang, Karol

2015-03-01

We present a brief overview of current plans to pursue two challenging goals, resolution of the neutrino mass hierarchy and determination of the CP phase of the PMNS neutrino mixing matrix. Future prospects include large atmospheric experiments, PINGU, ORCA, and INO-ICAL, medium baseline reactor experiments, JUNO and RENO- 50, and long baseline accelerator experiments, LBNE, LBNO, and Hyper-Kamiokande. There are also new initiatives emerging, ESSνSB at the European Spallation Source, and CHIPS in the NuMI neutrino beam. This is a multifaceted, vigorous, and technically difficult world-wide program. It will likely take more than a decade to start reaping its benefits.

LFR "Lead-Cooled Fast Reactor"

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

Cinotti, L; Fazio, C; Knebel, J

2006-05-11

The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small,more » transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected to benefit greatly from ongoing lead and lead-alloy technology development already being carried out in different institutes participating in this STREP. This is particularly true in Europe where a large R&D program associated with the development of Accelerator Driven Systems (ADS) is being actively pursued. The general objective of the ELSY project is to design an innovative lead-cooled fast reactor complemented by an analytical effort to assess the existing knowledge base in the field of lead-alloy coolants (i.e., lead-bismuth eutectic (LBE) and also lead/lithium) in order to extrapolate this knowledge base to pure lead. This analysis effort will be complemented with some limited R&D activities to acquire missing or confirmatory information about fundamental topics for ELSY that are not sufficiently covered in the ongoing European ADS program or elsewhere.« less

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

None, None

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

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

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

ERIC Educational Resources Information Center

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

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

78 FR 35056 - Effectiveness of the Reactor Oversight Process Baseline Inspection Program

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-11

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0125] Effectiveness of the Reactor Oversight Process... the effectiveness of the reactor oversight process (ROP) baseline inspection program with members of... Nuclear Reactor Regulations, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301...

10 CFR 72.26 - Contents of application: Technical specifications.

Code of Federal Regulations, 2011 CFR

2011-01-01

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

10 CFR 72.26 - Contents of application: Technical specifications.

Code of Federal Regulations, 2010 CFR

2010-01-01

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

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

PubMed

Wellock, Thomas R

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

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

Heavy-section steel technology and irradiation programs-retrospective and prospective views

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

Nanstad, Randy K; Bass, Bennett Richard; Rosseel, Thomas M

In 1965, the Atomic Energy Commission (AEC), at the advice of the Advisory Committee on Reactor Safeguards (ACRS), initiated the process that resulted in the establishment of the Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratory (ORNL). Dr. Spencer H. Bush of Battelle Northwest Laboratory, the man being honored by this symposium, representing the ACRS, was one of the Staff Advisors for the program and helped to guide its technical direction. In 1989, the Heavy-Section Steel Irradiation (HSSI) Program, formerly the HSST task on irradiation effects, was formed as a separate program, and this year the HSST/HSSImore » Programs, sponsored by the U.S. Nuclear Regulatory Commission (USNRC), celebrate 40 years of continuous research oriented toward the safety of light-water nuclear reactor pressure vessels. This paper presents a summary of results from those programs with a view to future activities. The HSST Program was established in 1967 and initially included extensive investigations of heavy-section low-alloy steel plates, forgings, and welds, including metallurgical studies, mechanical properties, fracture toughness (quasi-static and dynamic), fatigue crack-growth, and crack arrest toughness. Also included were irradiation effects studies, thermal shock analyses, testing of thick-section tensile and fracture specimens, and non-destructive testing. In the subsequent decades, the HSST Program conducted extensive large-scale experiments with intermediate-size vessels (with varying size flaws) pressurized to failure, similar experiments under conditions of thermal shock and even pressurized thermal shock (PTS), wide-plate crack arrest tests, and biaxial tests with cruciform-shaped specimens. Extensive analytical and numerical studies accompanied these experiments, including the development of computer codes such as the recent Fracture Analysis of Vessels Oak Ridge (FAVOR) code currently being used for PTS evaluations. In the absence of radiation damage to the RPV, fracture of the vessel is improbable. However, exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The HSSI Program has conducted a series of experiments to assess the effects of neutron irradiation on RPV material behavior, especially fracture toughness. These studies have included RPV plates and welds, varying chemical compositions, and fracture toughness specimens up to 4 in. thickness. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. Results from the HSST and HSSI Program are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety.« less

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

DOE PAGES

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

2015-07-17

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

Issues in electric power in India: Challenges and opportunities

NASA Astrophysics Data System (ADS)

Tongia, Rahul

This dissertation provides an examination of three facets of the Indian power program. The first issue we analyze is the current regulatory environment and guidelines in place for independent power producers and other generators, focusing on possible tradeoffs between prices and investor returns. The analysis shows that investor rates of return are significantly higher than the nominal 16% as stipulated by the Central Electricity Authority guidelines, and an uncertainty analysis reveals the relative importance of various input and project parameters. We discuss problems with the existing guidelines, and provide options for changes in policy. Adoption of modified guidelines that are more transparent and do not focus on project capital structures are likely to result in more affordable tariffs, less delays in project completion and yet provide adequate rates of return for investors. India's nuclear power program is based on indigenous materials and technology, with the potential for providing energy security for many decades. We examine the technical validity of this plan, especially the role of fast breeder reactors for extending the domestic uranium supplies. The analysis shows that breeding is unlikely to occur at anywhere near the rates envisioned, leading to a slow growth of fast breeder reactors. In addition, domestic uranium reserves restrict growth of Pressurized Heavy Water Reactors, which are likely to be the main contributors to nuclear capacity in the short term. To increase the share of nuclear power in the coming decades, India should consider the construction of a number of large thermal reactors based on indigenous and imported uranium. We also present policy options for such changes to India's nuclear power program. This dissertation examines in detail the policy, technology, and economics of an overland pipeline supplying natural gas to India and Pakistan. Such a pipeline would be shared by both countries, and would be a strong confidence building measure, offering a unique opportunity for cooperation. As natural gas pipelines exhibit significant economies of scale, a shared pipeline would also offer the lowest price natural gas for both countries. This study addresses some of the potential concerns, suggesting options for overcoming security of supply worries. (Abstract shortened by UMI.)

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

NASA Astrophysics Data System (ADS)

1980-08-01

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

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

Edgue, E.

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-21

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

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

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

Smith, Curtis

2013-09-01

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

The status of beryllium technology for fusion

NASA Astrophysics Data System (ADS)

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

2000-12-01

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

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

NASA Astrophysics Data System (ADS)

Gordon, Barry; Garcia, Susan

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

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

NASA Astrophysics Data System (ADS)

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

2010-11-01

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

Integrated low emissions cleanup system for direct coal-fueled turbines

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

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

1992-01-01

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

Integrated low emissions cleanup system for direct coal-fueled turbines

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

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

1992-12-31

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

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...

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

Mulder, R. U.; Benneche, P. E.; Hosticka, B.

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these users institutions is enhanced by the use of the nuclear facilities.

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

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

Nuclear reactors built, being built, or planned 1993

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

Not Available

1993-08-01

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

Small low mass advanced PBR's for propulsion

NASA Astrophysics Data System (ADS)

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

1993-10-01

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

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

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

Gilli, Ludivine; Charron, Sylvie

2012-07-01

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

DE-NE0008277_PROTEUS final technical report 2018

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

Enqvist, Andreas

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

Radiation Embrittlement Archive Project

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

Klasky, Hilda B; Bass, Bennett Richard; Williams, Paul T

2013-01-01

The Radiation Embrittlement Archive Project (REAP), which is being conducted by the Probabilistic Integrity Safety Assessment (PISA) Program at Oak Ridge National Laboratory under funding from the U.S. Nuclear Regulatory Commission s (NRC) Office of Nuclear Regulatory Research, aims to provide an archival source of information about the effect of neutron radiation on the properties of reactor pressure vessel (RPV) steels. Specifically, this project is an effort to create an Internet-accessible RPV steel embrittlement database. The project s website, https://reap.ornl.gov, provides information in two forms: (1) a document archive with surveillance capsule(s) reports and related technical reports, in PDF format,more » for the 104 commercial nuclear power plants (NPPs) in the United States, with similar reports from other countries; and (2) a relational database archive with detailed information extracted from the reports. The REAP project focuses on data collected from surveillance capsule programs for light-water moderated, nuclear power reactor vessels operated in the United States, including data on Charpy V-notch energy testing results, tensile properties, composition, exposure temperatures, neutron flux (rate of irradiation damage), and fluence, (Fast Neutron Fluence a cumulative measure of irradiation for E>1 MeV). Additionally, REAP contains data from surveillance programs conducted in other countries. REAP is presently being extended to focus on embrittlement data analysis, as well. This paper summarizes the current status of the REAP database and highlights opportunities to access the data and to participate in the project.« less

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

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

Kot, C. A.

1998-02-10

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

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

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.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-02-18

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

Fusion materials semiannual progress report for the period ending December 31, 1996

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

NONE

1997-04-01

This is the twenty-first in a series of semiannual technical progress reports on fusion materials. This report combines the full spectrum of research and development activities on both metallic and non-metallic materials with primary emphasis on the effects of the neutronic and chemical environment on the properties and performance of materials for in-vessel components. This effort forms one element of the materials program being conducted in support of the Fusion Energy Sciences Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reportedmore » separately. The report covers the following topics: vanadium alloys; silicon carbide composite materials; ferritic/martensitic steels; copper alloys and high heat flux materials; austenitic stainless steels; insulating ceramics and optical materials; solid breeding materials; radiation effects, mechanistic studies and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; and irradiation facilities, test matrices, and experimental methods.« less

Multiple model approach to evaluation of accelerated carbonation for steelmaking slag in a slurry reactor.

PubMed

Pan, Shu-Yuan; Liu, Hsing-Lu; Chang, E-E; Kim, Hyunook; Chen, Yi-Hung; Chiang, Pen-Chi

2016-07-01

Basic oxygen furnace slag (BOFS) exhibits highly alkaline properties due to its high calcium content, which is beneficial to carbonation reaction. In this study, accelerated carbonation of BOFS was evaluated under different reaction times, temperatures, and liquid-to-solid (L/S) ratios in a slurry reactor. CO2 mass balance within the slurry reactor was carried out to validate the technical feasibility of fixing gaseous CO2 into solid precipitates. After that, a multiple model approach, i.e., theoretical kinetics and empirical surface model, for carbonation reaction was presented to determine the maximal carbonation conversion of BOFS in a slurry reactor. On one hand, the reaction kinetics of BOFS carbonation was evaluated by the shrinking core model (SCM). Calcite (CaCO3) was identified as a reaction product through the scanning electronic microscopy and X-ray diffraction analyses, which provided the rationale of applying the SCM in this study. The rate-limiting step of carbonation was found to be ash-diffusion controlled, and the effective diffusivity for carbonation of BOFS in a slurry reactor were determined accordingly. On the other hand, the carbonation conversion of BOFS was predicted by the response surface methodology (RSM) via a nonlinear mathematical programming. According to the experimental data, the highest carbonation conversion of BOFS achieved was 57% under an L/S ratio of 20 mL g(-1), a CO2 flow rate of 0.1 L min(-1), and a pressure of 101.3 kPa at 50 °C for 120 min. Furthermore, the applications and limitations of SCM and RSM were examined and exemplified by the carbonation of steelmaking slags. Copyright © 2016 Elsevier Ltd. All rights reserved.

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.

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

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

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

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

NASA Technical Reports Server (NTRS)

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

2001-01-01

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

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

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

Yacout, A. M.; Billone, M. C.

2016-09-16

The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of datamore » were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.« less

Expanded scope of training and education programs at the UFTR

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

Vernetson, W.G.; Whaley, P.M.

1985-01-01

Historically, the University of Florida Training Reactor (UFTR) has been used to train both hot and cold license reactor operator candidates in intensive two- and three-week training programs consisting of a correlated set of classroom lectures, hands-on reactor operations, and laboratory exercises. These training programs provide nuclear plant operating staff with fundamental operational experience in understanding, controlling, and evaluating subcritical multiplication, reactivity effects, reactivity manipulations, and reactor operations; a sufficient number of startups and shutdowns is also assured. The UDTR is also used in a nuclear engineering course entitled ''Principles of Nuclear Reactor Operations.'' The purpose of this paper ismore » to report the results of efforts to redirect and refine tractor operations educational and training programs at the UFTR.« less

Dismantling the nuclear research reactor Thetis

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

Michiels, P.

The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3 m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storagemore » rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were used, which must be removed in controlled conditions. The ventilation system is considered free from nuclear contamination but it contains asbestos. This paper covers the organization of the dismantling work, the technical execution aspect and conclusions already known (dismantling is ongoing as this is written). (authors)« less

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

Lunar base thermoelectric power station study

NASA Technical Reports Server (NTRS)

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

2006-01-01

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

Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

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

NONE

1994-04-30

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readilymore » achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.« less

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

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

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

2008-07-15

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

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

NASA Technical Reports Server (NTRS)

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

2008-01-01

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

A process economic assessment of hydrocarbon biofuels production using chemoautotrophic organisms

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

Khan, NE; Myers, JA; Tuerk, AL

Economic analysis of an ARPA-e Electrofuels (http://arpa-e.energy.gov/?q=arpa-e-programs/electrofuels) process is presented, utilizing metabolically engineered Rhodobacter capsulatus or Ralstonia eutropha to produce the C30+ hydrocarbon fuel, botryococcene, from hydrogen, carbon dioxide, and oxygen. The analysis is based on an Aspen plus (R) bioreactor model taking into account experimentally determined Rba. capsulatus and Rls. eutropha growth and maintenance requirements, reactor residence time, correlations for gas-liquid mass-transfer coefficient, gas composition, and specific cellular fuel productivity. Based on reactor simulation results encompassing technically relevant parameter ranges, the capital and operating costs of the process were estimated for 5000 bbl-fuel/day plant and used to predict fuelmore » cost. Under the assumptions used in this analysis and crude oil prices, the Levelized Cost of Electricity (LCOE) required for economic feasibility must be less than 2(sic)/kWh. While not feasible under current market prices and costs, this work identifies key variables impacting process cost and discusses potential alternative paths toward economic feasibility. (C) 2014 Elsevier Ltd. All rights reserved.« less

Supervisory Control System Architecture for Advanced Small Modular Reactors

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

Cetiner, Sacit M; Cole, Daniel L; Fugate, David L

2013-08-01

This technical report was generated as a product of the Supervisory Control for Multi-Modular SMR Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (SMR) Research and Development Program of the U.S. Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular advanced SMR (AdvSMR) plants. This research activity advances the state-of-the art by incorporating decision making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides a brief history ofmore » hierarchical functional architectures and the current state-of-the-art, describes a reference AdvSMR to show the dependencies between systems, presents a hierarchical structure for supervisory control, indicates the importance of understanding trip setpoints, applies a new theoretic approach for comparing architectures, identifies cyber security controls that should be addressed early in system design, and describes ongoing work to develop system requirements and hardware/software configurations.« less

Thermionic fuel element for the S-prime reactor

NASA Astrophysics Data System (ADS)

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

1993-01-01

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

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

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

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

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

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

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

Development of the reactor antineutrino detection technology within the iDream project

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Kinetics of Supercritical Water Oxidation

DTIC Science & Technology

1995-12-31

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

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

NASA Astrophysics Data System (ADS)

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

2014-08-01

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

Nuclear reactors built, being built, or planned 1996

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

NONE

1997-08-01

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

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

NASA Astrophysics Data System (ADS)

Stork, D.; Zinkle, S. J.

2017-09-01

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

Biofuels Refining Engineering

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

Lobban, Lance

The goal of this project is the development of novel catalysts and knowledge of reaction pathways and mechanisms for conversion of biomass-based compounds to fuels that are compatible with oil-based fuels and with acceptable or superior fuel properties. The research scope included both catalysts to convert lignocellulosic biomass-based molecules (from pyrolysis) and vegetable oil-based molecules (i.e., triglycerides and fatty acid methyl esters). This project comprised five technical tasks. Each task is briefly introduced below, and major technical accomplishments summarized. Technical accomplishments were described in greater detail in the quarterly progress reports, and in even more detail in the >50 publicationsmore » acknowledging this DoE project funding (list of publications and presentations included at the end of this report). The results of this research added greatly to the knowledge base necessary for upgrading of pyrolysis oil to hydrocarbon fuels and chemicals, and for conversion of vegetable oils to fungible diesel fuel. Numerous new catalysts and catalytic reaction systems were developed for upgrading particular compounds or compound families found in the biomass-based pyrolysis oils and vegetable oils. Methods to mitigate catalyst deactivation were investigated, including novel reaction/separation systems. Performance and emission characteristics of biofuels in flames and engines were measured. Importantly, the knowledge developed from this project became the basis for a subsequent collaborative proposal led by our research group, involving researchers from the University of Wisconsin, the University of Pittsburg, and the Idaho National Lab, for the DoE Carbon, Hydrogen and Separations Efficiency (CHASE) program, which was subsequently funded (one of only four projects awarded in the CHASE program). The CHASE project examined novel catalytic processes for lignocellulosic biomass conversion as well as technoeconomic analyses for process options for maximum carbon capture and hydrogen efficiency. Our research approach combined catalyst synthesis, measurements of catalyst activity and selectivity in different reactor systems and conditions, and detailed catalyst characterization to develop fundamental understanding of reaction pathways and the capability to predict product distributions. Nearly all of the candidate catalysts were prepared in-house via standard techniques such as impregnation, co-impregnation, or chemical vapor deposition. Supports were usually purchased, but in some cases coprecipitation was used to simultaneously create the support and active component, which can be advantageous for strong active component-support interactions and for achieving high active component dispersion. In-house synthesis also allowed for studies of the effects on catalyst activity and selectivity of such factors as support porosity, calcination temperature, and reduction/activation conditions. Depending on the physical characteristics of the molecule, catalyst activity measurements were carried out in tubular flow reactors (for vapor phase reactions) or stirred tank reactors (for liquid phase reactions) over a wide range of pressures and temperatures. Reactant and product concentrations were measured using gas chromatography (both on-line and off-line, with TCD, FID, and/or mass spectrometric detection). For promising catalysts, detailed physicochemical characterization was carried out using FTIR, Raman, XPS, and XRD spectroscopies (all available in our laboratories) and TEM spectroscopy (available at OU). Additional methods included temperature programmed techniques (TPD, TPO) and surface area measurements by nitrogen adsorption techniques.« less

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

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

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

2010-11-01

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

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

Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

Worldwide activities related to nuclear fuel cycle and radioactive waste management programs are summarized. Several trends have developed in waste management strategy: All countries having to dispose of reprocessing wastes plan on conversion of the high-level waste (HLW) stream to a borosilicate glass and eventual emplacement of the glass logs, suitably packaged, in a deep geologic repository. Countries that must deal with plutonium-contaminated waste emphasize pluonium recovery, volume reduction and fixation in cement or bitumen in their treatment plans and expect to use deep geologic repositories for final disposal. Commercially available, classical engineering processing are being used worldwide to treatmore » and immobilize low- and intermediate-level wastes (LLW, ILW); disposal to surface structures, shallow-land burial and deep-underground repositories, such as played-out mines, is being done widely with no obvious technical problems. Many countries have established extensive programs to prepare for construction and operation of geologic repositories. Geologic media being studied fall into three main classes: argillites (clay or shale); crystalline rock (granite, basalt, gneiss or gabbro); and evaporates (salt formations). Most nations plan to allow 30 years or longer between discharge of fuel from the reactor and emplacement of HLW or spent fuel is a repository to permit thermal and radioactive decay. Most repository designs are based on the mined-gallery concept, placing waste or spent fuel packages into shallow holes in the floor of the gallery. Many countries have established extensive and costly programs of site evaluation, repository development and safety assessment. Two other waste management problems are the subject of major R and D programs in several countries: stabilization of uranium mill tailing piles; and immobilization or disposal of contaminated nuclear facilities, namely reactors, fuel cycle plants and R and D laboratories.« less

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

NONE

The High Ranking Facilities Deactivation Project (HRFDP), commissioned by the US Department of Energy Nuclear Materials and Facility Stabilization Program, is to place four primary high-risk surplus facilities with 28 associated ancillary facilities at Oak Ridge National Laboratory in a safe, stable, and environmentally sound condition as rapidly and economically as possible. The facilities will be deactivated and left in a condition suitable for an extended period of minimized surveillance and maintenance (S and M) prior to decontaminating and decommissioning (D and D). These four facilities include two reactor facilities containing spent fuel. One of these reactor facilities also containsmore » 55 tons of sodium with approximately 34 tons containing activated sodium-22, 2.5 tons of lithium hydride, approximately 100 tons of potentially contaminated lead, and several other hazardous materials as well as bulk quantities of contaminated scrap metals. The other two facilities to be transferred include a facility with a bank of hot cells containing high levels of transferable contamination and also a facility containing significant quantities of uranyl nitrate and quantities of transferable contamination. This work plan documents the objectives, technical requirements, and detailed work plans--including preliminary schedules, milestones, and conceptual FY 1996 cost estimates--for the Oak Ridge National Laboratory (ORNL). This plan has been developed by the Environmental Restoration (ER) Program of Lockheed Martin Energy Systems (Energy Systems) for the US Department of Energy (DOE) Oak Ridge Operations Office (ORO).« less

CHARACTERISTIC QUALITIES OF SOME ATOMIC POWER STATIONS (in Hungarian)

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

Ligeti, G.

1962-04-01

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

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

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

Piani, C.S.B.

2008-07-15

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

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

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

Sullivan, Edmund J.; Anderson, Michael T.

2014-06-10

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

N Reactor Deactivation Program Plan. Revision 4

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

Walsh, J.L.

1993-12-01

This N Reactor Deactivation Program Plan is structured to provide the basic methodology required to place N Reactor and supporting facilities {center_dot} in a radiologically and environmentally safe condition such that they can be decommissioned at a later date. Deactivation will be in accordance with facility transfer criteria specified in Department of Energy (DOE) and Westinghouse Hanford Company (WHC) guidance. Transition activities primarily involve shutdown and isolation of operational systems and buildings, radiological/hazardous waste cleanup, N Fuel Basin stabilization and environmental stabilization of the facilities. The N Reactor Deactivation Program covers the period FY 1992 through FY 1997. The directivemore » to cease N Reactor preservation and prepare for decommissioning was issued by DOE to WHC on September 20, 1991. The work year and budget data supporting the Work Breakdown Structure in this document are found in the Activity Data Sheets (ADS) and the Environmental Restoration Program Baseline, that are prepared annually.« less

Experiences in utilization of research reactors in Yugoslavia

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

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

1971-06-15

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

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

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

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

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.

F Reactor Inspection

ScienceCinema

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2018-01-16

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

F Reactor Inspection

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosuremore » and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."« less

Edwin I. Hatch nuclear plant implementation of improved technical specifications

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

Mahler, S.R.; Pendry, D.

1994-12-31

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

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

NASA Astrophysics Data System (ADS)

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

2014-02-01

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

Scoping Calculations of Power Sources for Nuclear Electric Propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1994-01-01

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

Astronaut William Anders Official Portrait

NASA Technical Reports Server (NTRS)

1967-01-01

This is the official NASA portrait of astronaut William Anders. Anders was commissioned in the air Force after graduation from the Naval Academy and served as a fighter pilot in all-weather interception squadrons of the Air Defense Command. Later he was responsible for technical management of nuclear power reactor shielding and radiation effects programs while at the Air Force Weapons Laboratory in New Mexico. In 1964, Anders was selected by the National Aeronautics and Space Administration (NASA) as an astronaut with responsibilities for dosimetry, radiation effects and environmental controls. He was backup pilot for the Gemini XI, Apollo 11 flights, and served as lunar module (LM) pilot for Apollo 8, the first lunar orbit mission in December 1968. He has logged more than 6,000 hours flying time.

Preliminary Framework for Human-Automation Collaboration

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

Oxstrand, Johanna Helene; Le Blanc, Katya Lee; Spielman, Zachary Alexander

The Department of Energy’s Advanced Reactor Technologies Program sponsors research, development and deployment activities through its Next Generation Nuclear Plant, Advanced Reactor Concepts, and Advanced Small Modular Reactor (aSMR) Programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. The Human Automation Collaboration (HAC) Research Project is located under the aSMR Program, which identifies developing advanced instrumentation and controls and human-machine interfaces as one of four key research areas. It is expected that the new nuclear power plant designs will employ technology significantly more advanced than the analog systems in the existing reactor fleetmore » as well as utilizing automation to a greater extent. Moving towards more advanced technology and more automation does not necessary imply more efficient and safer operation of the plant. Instead, a number of concerns about how these technologies will affect human performance and the overall safety of the plant need to be addressed. More specifically, it is important to investigate how the operator and the automation work as a team to ensure effective and safe plant operation, also known as the human-automation collaboration (HAC). The focus of the HAC research is to understand how various characteristics of automation (such as its reliability, processes, and modes) effect an operator’s use and awareness of plant conditions. In other words, the research team investigates how to best design the collaboration between the operators and the automated systems in a manner that has the greatest positive impact on overall plant performance and reliability. This report addresses the Department of Energy milestone M4AT-15IN2302054, Complete Preliminary Framework for Human-Automation Collaboration, by discussing the two phased development of a preliminary HAC framework. The framework developed in the first phase was used as the basis for selecting topics to be investigated in more detail. The results and insights gained from the in-depth studies conducted during the second phase were used to revise the framework. This report describes the basis for the framework developed in phase 1, the changes made to the framework in phase 2, and the basis for the changes. Additional research needs are identified and presented in the last section of the report.« less

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

PubMed

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

2006-04-01

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

Code qualification of structural materials for AFCI advanced recycling reactors.

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

Natesan, K.; Li, M.; Majumdar, S.

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Codemore » Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.« less

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

DOE PAGES

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

2017-03-26

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

Environmental Cracking and Irradiation Resistant Stainless Steels by Additive Manufacturing

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

Rebak, Raul B.; Lou, Xiaoyuan

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

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

NASA Technical Reports Server (NTRS)

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

1985-01-01

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

Reactor engineering support of operations at the Davis-Besse nuclear power station

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

Kelley, D.B.

1995-12-31

Reactor engineering functions differ greatly from unit to unit; however, direct support of the reactor operators during reactor startups and operational transients is common to all units. This paper summarizes the support the reactor engineers provide the reactor operators during reactor startups and power changes through the use of automated computer programs at the Davis-Besse nuclear power station.

25 CFR 170.161 - What is the Indian Local Technical Assistance Program?

Code of Federal Regulations, 2010 CFR

2010-04-01

... 25 Indians 1 2010-04-01 2010-04-01 false What is the Indian Local Technical Assistance Program... Technical Assistance Program § 170.161 What is the Indian Local Technical Assistance Program? The Indian Local Technical Assistance Program (Indian LTAP) is authorized under 23 U.S.C. 504(b), and §§ 170.161...

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

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

Mouradian, E.M.

1966-02-16

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

ECUT (Energy Conversion and Utilization Technologies) program: Biocatalysis project

NASA Technical Reports Server (NTRS)

Baresi, Larry

1989-01-01

The Annual Report presents the fiscal year (FY) 1988 research activities and accomplishments, for the Biocatalysis Project of the U.S. Department of Energy, Energy Conversion and Utilization Technologies (ECUT) Division. The ECUT Biocatalysis Project is managed by the Jet Propulsion Laboratory, California Institute of Technology. The Biocatalysis Project is a mission-oriented, applied research and exploratory development activity directed toward resolution of the major generic technical barriers that impede the development of biologically catalyzed commercial chemical production. The approach toward achieving project objectives involves an integrated participation of universities, industrial companies and government research laboratories. The Project's technical activities were organized into three work elements: (1) The Molecular Modeling and Applied Genetics work element includes research on modeling of biological systems, developing rigorous methods for the prediction of three-dimensional (tertiary) protein structure from the amino acid sequence (primary structure) for designing new biocatalysis, defining kinetic models of biocatalyst reactivity, and developing genetically engineered solutions to the generic technical barriers that preclude widespread application of biocatalysis. (2) The Bioprocess Engineering work element supports efforts in novel bioreactor concepts that are likely to lead to substantially higher levels of reactor productivity, product yields and lower separation energetics. Results of work within this work element will be used to establish the technical feasibility of critical bioprocess monitoring and control subsystems. (3) The Bioprocess Design and Assessment work element attempts to develop procedures (via user-friendly computer software) for assessing the energy-economics of biocatalyzed chemical production processes, and initiation of technology transfer for advanced bioprocesses.

ECUT (Energy Conversion and Utilization Technologies) program: Biocatalysis project

NASA Astrophysics Data System (ADS)

Baresi, Larry

1989-03-01

The Annual Report presents the fiscal year (FY) 1988 research activities and accomplishments, for the Biocatalysis Project of the U.S. Department of Energy, Energy Conversion and Utilization Technologies (ECUT) Division. The ECUT Biocatalysis Project is managed by the Jet Propulsion Laboratory, California Institute of Technology. The Biocatalysis Project is a mission-oriented, applied research and exploratory development activity directed toward resolution of the major generic technical barriers that impede the development of biologically catalyzed commercial chemical production. The approach toward achieving project objectives involves an integrated participation of universities, industrial companies and government research laboratories. The Project's technical activities were organized into three work elements: (1) The Molecular Modeling and Applied Genetics work element includes research on modeling of biological systems, developing rigorous methods for the prediction of three-dimensional (tertiary) protein structure from the amino acid sequence (primary structure) for designing new biocatalysis, defining kinetic models of biocatalyst reactivity, and developing genetically engineered solutions to the generic technical barriers that preclude widespread application of biocatalysis. (2) The Bioprocess Engineering work element supports efforts in novel bioreactor concepts that are likely to lead to substantially higher levels of reactor productivity, product yields and lower separation energetics. Results of work within this work element will be used to establish the technical feasibility of critical bioprocess monitoring and control subsystems. (3) The Bioprocess Design and Assessment work element attempts to develop procedures (via user-friendly computer software) for assessing the energy-economics of biocatalyzed chemical production processes, and initiation of technology transfer for advanced bioprocesses.

U.S. Department of Energy physical protection upgrades at the Latvian Academy of Sciences Nuclear Research Center, Latvia

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

Haase, M.; Hine, C.; Robertson, C.

1996-12-31

Approximately five years ago, the Safe, Secure Dismantlement program was started between the US and countries of the Former Soviet Union (FSU). The purpose of the program is to accelerate progress toward reducing the risk of nuclear weapons proliferation, including such threats as theft, diversion, and unauthorized possession of nuclear materials. This would be accomplished by strengthening the material protection, control, and accounting systems within the FSU countries. Under the US Department of Energy`s program of providing cooperative assistance to the FSU countries in the areas of Material Protection, Control, and Accounting (MPC and A), the Latvian Academy of Sciencesmore » Nuclear Research Center (LNRC) near Riga, Latvia, was identified as a candidate site for a cooperative MPC and A project. The LNRC is the site of a 5-megawatt IRT-C pool-type research reactor. This paper describes: the process involved, from initial contracting to project completion, for the physical protection upgrades now in place at the LNRC; the intervening activities; and a brief overview of the technical aspects of the upgrades.« less

Nuclear waste disposal in subseabed geologic formatons: the Seabed Disposal Program

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

Anderson, D.R.

1979-05-01

The goal of the Seabed Disposal Program is to assess the technical and environmental feasibility of using geologic formations under the sea floor for the disposal of processed high-level radioactive wastes or repackaged spent reactor fuel. Studies are focused on the abyssal hill regions of the sea floors in the middle of tectonic plates and under massive surface current gyres. The red-clay sediments here are from 50 to 100 meters thick, are continuously depositional (without periods of erosion), and have been geologically and climatologically stable for millions of years. Mineral deposits and biological activity are minimal, and bottom currents aremore » weak and variable. Five years of research have revealed no technological reason why nuclear waste disposal in these areas would be impractical. However, scientific assessment is not complete. Also, legal political, and sociological factors may well become the governing elements in such use of international waters. These factors are being examined as part of the work of the Seabed Working Group, an international adjunct of the Seabed Program, with members from France, England, Japan, Canada, and the United States.« 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

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

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

Cottrell, W.B.; Klein, A.

1977-02-23

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

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

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

Block, R.C.; Feiner, F.

1995-09-01

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

Automatic reactor model synthesis with genetic programming.

PubMed

Dürrenmatt, David J; Gujer, Willi

2012-01-01

Successful modeling of wastewater treatment plant (WWTP) processes requires an accurate description of the plant hydraulics. Common methods such as tracer experiments are difficult and costly and thus have limited applicability in practice; engineers are often forced to rely on their experience only. An implementation of grammar-based genetic programming with an encoding to represent hydraulic reactor models as program trees should fill this gap: The encoding enables the algorithm to construct arbitrary reactor models compatible with common software used for WWTP modeling by linking building blocks, such as continuous stirred-tank reactors. Discharge measurements and influent and effluent concentrations are the only required inputs. As shown in a synthetic example, the technique can be used to identify a set of reactor models that perform equally well. Instead of being guided by experience, the most suitable model can now be chosen by the engineer from the set. In a second example, temperature measurements at the influent and effluent of a primary clarifier are used to generate a reactor model. A virtual tracer experiment performed on the reactor model has good agreement with a tracer experiment performed on-site.

75 FR 13521 - Centers for Independent Living Program-Training and Technical Assistance

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-22

... DEPARTMENT OF EDUCATION Centers for Independent Living Program--Training and Technical Assistance... for Independent Living Program--Training and Technical Assistance (CIL-TA program). The Assistant... appropriated for the CIL program to provide training and technical assistance to CILs, agencies eligible to...

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

PubMed

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

2008-05-01

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

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

PubMed

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

2018-02-19

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

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

NASA Astrophysics Data System (ADS)

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

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

Integrated intelligent systems in advanced reactor control rooms

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

Beckmeyer, R.R.

1989-01-01

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

Degradation of toxaphene in water during anaerobic and aerobic conditions.

PubMed

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

2004-08-01

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

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

NASA Astrophysics Data System (ADS)

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

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

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

NASA Astrophysics Data System (ADS)

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

2017-11-01

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

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

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

None

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

Production assurance program strategy for N Reactor balance of plant systems

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

House, R.D.; Bitten, E.J.; Keenan, J.P.

1986-03-18

A production assurance program has been established for N Reactor, a dual purpose reactor plant, operated to produce special nuclear materials and steam for electricity. N Reactor, which began operation in December 1963, is now approaching the end of its design life. This paper describes the two phase program for Balance of Plant (BOP) systems. The Phase I evaluation has been completed and indications are that the lifetime of systems and components could be extended by implementing appropriate surveillance, operations and maintenance strategies. In Phase II, a thorough evaluation of components and systems is underway and action items are beingmore » identified which will allow component and system extended operation.« less

Integrated Ceramic Membrane System for Hydrogen Production

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

Schwartz, Joseph; Lim, Hankwon; Drnevich, Raymond

2010-08-05

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

The North Korean nuclear dilemma.

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

Hecker, Siegfried S.

2004-01-01

The current nuclear crisis, the second one in ten years, erupted when North Korea expelled international nuclear inspectors in December 2002, then withdrew from the Nuclear Nonproliferation Treaty (NPT), and claimed to be building more nuclear weapons with the plutonium extracted from the spent fuel rods heretofore stored under international inspection. These actions were triggered by a disagreement over U.S. assertions that North Korea had violated the Agreed Framework (which froze the plutonium path to nuclear weapons to end the first crisis in 1994) by clandestinely developing uranium enrichment capabilities providing an alternative path to nuclear weapons. With Stanford Universitymore » Professor John Lewis and three other Americans, I was allowed to visit the Yongbyon Nuclear Center on Jan. 8, 2004. We toured the 5 MWe reactor, the 50 MWe reactor construction site, the spent fuel pool storage building, and the radiochemical laboratory. We concluded that North Korea has restarted its 5 MWe reactor (which produces roughly 6 kg of plutonium annually), it removed the 8000 spent fuel rods that were previously stored under IAEA safeguards from the spent fuel pool, and that it most likely extracted the 25 to 30 kg of plutonium contained in these fuel rods. Although North Korean officials showed us what they claimed was their plutonium metal product from this reprocessing campaign, we were not able to conclude definitively that it was in fact plutonium metal and that it came from the most recent reprocessing campaign. Nevertheless, our North Korean hosts demonstrated that they had the capability, the facility and requisite capacity, and the technical expertise to produce plutonium metal. On the basis of our visit, we were not able to address the issue of whether or not North Korea had a 'deterrent' as claimed - that is, we were not able to conclude that North Korea can build a nuclear device and that it can integrate nuclear devices into suitable delivery systems. However, based on the capabilities we saw, we must assume that North Korea has the capability to produce a crude nuclear device. On the matter of uranium enrichment programs, our host categorically denied that North Korea has a uranium enrichment program - he said, 'we have no program, no equipment, and no technical expertise for uranium enrichment.' The denials were not convincing at the time and since then have proven to be quite hollow by the revelations of A.Q. Khan's nuclear black market activities. There is no easy solution to the nuclear crisis in North Korea. A military strike to eliminate the nuclear facilities was never very attractive and now has been overcome by events. The principal threat is posed by a stockpile of nuclear weapons and weapons-grade plutonium. We have no way of finding where either may be hidden. A diplomatic solution remains the only path forward, but it has proven elusive. All sides have proclaimed a nuclear weapons-free Korean Peninsula as the end goal. The U.S. Government has chosen to negotiate with North Korea by means of the six-party talks. It has very clearly outlined its position of insisting on complete, verifiable, irreversible dismantlement of all North Korean nuclear programs. North Korea has offered several versions of 're-freezing' its plutonium program while still denying a uranium enrichment program. It has insisted on simultaneous and reciprocal steps to a final solution. Regardless of which diplomatic path is chosen, the scientific challenges of eliminating the North Korean nuclear weapons programs (and its associated infrastructure) in a safe, secure, and verifiable manner are immense. The North Korean program is considerably more complex and developed than the fledgling Iraqi program of 1991 and Libyan program of 2004. It is more along the lines, but more complex than that of South Africa in the early 1990s. Actions taken or not taken by the North Koreans at their nuclear facilities during the course of the ongoing diplomatic discussions are key to whether or not the nuclear program can be eliminated safely and securely, and they will greatly influence the price tag for such operations. Moreover, they will determine whether or not one can verify complete elimination. Hence, cooperation of the North Koreans now and during the dismantlement and elimination stages is crucial. Technical discussions among specialists, perhaps within the framework of the working groups of the six-party talks, could be very productive in setting the stage for an effective, verifiable elimination of North Korea's nuclear weapons program.« less

Safety philosophy of gas turbine high temperature reactor (GTHTR300)

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

Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

2002-07-01

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

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

NASA Technical Reports Server (NTRS)

Sibille, Laurent; Dominguez, Jesus A.

2012-01-01

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

Reactor concepts for bioelectrochemical syntheses and energy conversion.

PubMed

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

2014-12-01

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

Design of virtual SCADA simulation system for pressurized water reactor

NASA Astrophysics Data System (ADS)

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

2016-02-01

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

Proceedings of the 1994 international meeting on reduced enrichment for research and test reactors

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

NONE

1997-08-01

This meeting brought together participants in the international effort to minimize and eventually eliminate the use of highly enriched uranium in civilian nuclear programs. Papers cover the following topics: National programs; fuel cycle; nuclear fuels; analyses; advanced reactors; and reactor conversions. Selected papers have been indexed separately for inclusion to the Energy Science and Technology Database.

Cyber security evaluation of II&C technologies

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

Thomas, Ken

The Light Water Reactor Sustainability (LWRS) Program is a research and development program sponsored by the Department of Energy, which is conducted in close collaboration with industry to provide the technical foundations for licensing and managing the long-term, safe and economical operation of current nuclear power plants The LWRS Program serves to help the US nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Within the LWRS Program, the Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway conducts targeted research and development (R&D) tomore » address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. The II&C Pathway is conducted by Idaho National Laboratory (INL). Cyber security is a common concern among nuclear utilities and other nuclear industry stakeholders regarding the digital technologies that are being developed under this program. This concern extends to the point of calling into question whether these types of technologies could ever be deployed in nuclear plants given the possibility that the information in them can be compromised and the technologies themselves can potentially be exploited to serve as attack vectors for adversaries. To this end, a cyber security evaluation has been conducted of these technologies to determine whether they constitute a threat beyond what the nuclear plants already manage within their regulatory-required cyber security programs. Specifically, the evaluation is based on NEI 08-09, which is the industry’s template for cyber security programs and evaluations, accepted by the Nuclear Regulatory Commission (NRC) as responsive to the requirements of the nuclear power plant cyber security regulation found in 10 CFR 73.54. The evaluation was conducted by a cyber security team with expertise in nuclear utility cyber security programs and experience in conducting these evaluations. The evaluation has determined that, for the most part, cyber security will not be a limiting factor in the application of these technologies to nuclear power plant applications.« less

Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: Parts 1 and 2. Final report

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

Balkey, K.; Witt, F.J.; Bishop, B.A.

1995-06-01

Significant attention has been focused on the issue of reactor vessel pressurized thermal shock (PTS) for many years. Pressurized thermal shock transient events are characterized by a rapid cooldown at potentially high pressure levels that could lead to a reactor vessel integrity concern for some pressurized water reactors. As a result of regulatory and industry efforts in the early 1980`s, a probabilistic risk assessment methodology has been established to address this concern. Probabilistic fracture mechanics analyses are performed as part of this methodology to determine conditional probability of significant flaw extension for given pressurized thermal shock events. While recent industrymore » efforts are underway to benchmark probabilistic fracture mechanics computer codes that are currently used by the nuclear industry, Part I of this report describes the comparison of two independent computer codes used at the time of the development of the original U.S. Nuclear Regulatory Commission (NRC) pressurized thermal shock rule. The work that was originally performed in 1982 and 1983 to compare the U.S. NRC - VISA and Westinghouse (W) - PFM computer codes has been documented and is provided in Part I of this report. Part II of this report describes the results of more recent industry efforts to benchmark PFM computer codes used by the nuclear industry. This study was conducted as part of the USNRC-EPRI Coordinated Research Program for reviewing the technical basis for pressurized thermal shock (PTS) analyses of the reactor pressure vessel. The work focused on the probabilistic fracture mechanics (PFM) analysis codes and methods used to perform the PTS calculations. An in-depth review of the methodologies was performed to verify the accuracy and adequacy of the various different codes. The review was structured around a series of benchmark sample problems to provide a specific context for discussion and examination of the fracture mechanics methodology.« less

HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio

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

Bess, John D.

2014-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

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

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

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

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

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

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

Forsbacka, M.; Moore, M.

1989-12-01

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

NPR design basis

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

Locke, G.L.

1958-09-08

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

Technical Bases to Aid in the Decision of Conducting Full Power Ground Nuclear Tests for Space Fission Reactors

NASA Astrophysics Data System (ADS)

Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

2004-02-01

The extent to which, if any, full power ground nuclear testing of space reactors should be performed has been a point of discussion within the industry for decades. Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Is the test article an accurate representation of the flight system? Are the costs too restrictive? The obvious benefits of full power ground nuclear testing; obtaining systems integrated reliability data on a full-scale, complete end-to-end system; come at some programmatic risk. Safety related information is not obtained from a full-power ground nuclear test. This paper will discuss and assess these and other technical considerations essential in the decision to conduct full power ground nuclear-or alternative-tests.

The Technical Assistance Program: A Program Plan.

DTIC Science & Technology

1985-09-01

interests, talents, and hobbies of current employees who wish to become involved in the technical assistance program . 4. Capitalize on the corporate...Center San Diego, CA 92152 Telephone (619) 225-6281 B-33 TAP ( EMPLOYEE ) QUESTIONNAIRE - The Technical Assistance Program (TAP) links technically skilled...Brochure, "Retired Technologist" C-11 C. Sample Letter to NOSC Employees C-13 Brochure D. The Technical Assistance Program C-15 C-1

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

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.

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

DTIC Science & Technology

2009-10-28

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

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

DTIC Science & Technology

2009-07-17

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

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

DTIC Science & Technology

2009-12-23

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

Analytical methods in the high conversion reactor core design

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

Zeggel, W.; Oldekop, W.; Axmann, J.K.

High conversion reactor (HCR) design methods have been used at the Technical University of Braunschweig (TUBS) with the technological support of Kraftwerk Union (KWU). The present state and objectives of this cooperation between KWU and TUBS in the field of HCRs have been described using existing design models and current activities aimed at further development and validation of the codes. The hard physical and thermal-hydraulic boundary conditions of pressurized water reactor (PWR) cores with a high degree of fuel utilization result from the tight packing of the HCR fuel rods and the high fissionable plutonium content of the fuel. Inmore » terms of design, the problem will be solved with rod bundles whose fuel rods are adjusted by helical spacers to the proposed small rod pitches. These HCR properties require novel computational models for neutron physics, thermal hydraulics, and fuel rod design. By means of a survey of the codes, the analytical procedure for present-day HCR core design is presented. The design programs are currently under intensive development, as design tools with a solid, scientific foundation and with essential parameters that are widely valid and are required for a promising optimization of the HCR core. Design results and a survey of future HCR development are given. In this connection, the reoptimization of the PWR core in the direction of an HCR is considered a fascinating scientific task, with respect to both economic and safety aspects.« less

1170 MW/sub t/ HTGR steamer cogeneration plant: design and cost study

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

None

A conceptual design and cost study is presented for intermediate size high temperature gas-cooled reactor (HTGR) for industrial energy applications performed by United Engineers and Constructors Inc., (UE and C) and The General Atomic Company (GAC). The study is part of a program at ORNL and has the objective to provide support in the evaluation of the technical and economic feasibility of a single unit 1170 MW/sub t/ HTGR steam cycle cogeneration plant (referred to as the Steamer plant) for the production of industrial process energy. Inherent in the achievement of this objective, it was essential to perform a numbermore » of basic tasks such as the development of plant concept, capital cost estimate, project schedule and annual operation and maintenance (O and M) cost.« less

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

Bess, John D.; Sterbentz, James W.; Snoj, Luka

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Nuclear reactors built, being built, or planned, 1994

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

NONE

1995-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristicmore » and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

Nuclear reactors built, being built, or planned: 1995

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

NONE

1996-08-01

This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristicmore » and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

What You Should Get from a Professionally Oriented Master's Degree Program in Technical Communication.

ERIC Educational Resources Information Center

Carliner, Saul

1992-01-01

Cites reasons for pursuing a curriculum in technical communication, lists objectives a program should achieve, and outlines a four-part program that includes theory, professional skills, technical proficiency, and an internship. Lists schools offering programs in technical communication. (SR)

Removal of Legacy Low-Level Waste Reactor Moderator De-ionizer Resins Highly Contaminated with Carbon-14 from the 'Waste with no Path to Disposal List' Through Innovative Technical Analysis and Performance Assessment Techniques

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

Goldston, W.T.; Hiergesell, R.A.; Kaplan, D.I.

2006-07-01

At the Savannah River Site (SRS), nuclear production reactors used de-ionizers to control the chemistry of the reactor moderator during their operation to produce nuclear materials primarily for the weapons program. These de-ionizers were removed from the reactors and stored as a legacy waste and due to the relatively high carbon-14 (C-14) contamination (i.e., on the order of 740 giga becquerel (GBq) (20 curies) per de-ionizer) were considered a legacy 'waste with no path to disposal'. Considerable progress has been made in consideration of a disposal path for the legacy reactor de-ionizers. Presently, 48 - 50 de-ionizers being stored atmore » SRS have 'no path to disposal' because the disposal limit for C-14 in the SRS's low-level waste disposal facility's Intermediate Level Vault (ILV) is only 160 GBq (4.2 curies) per vault. The current C-14 ILV disposal limit is based on a very conservative analysis of the air pathway. The paper will describe the alternatives that were investigated that resulted in the selection of a route to pursue. This paper will then describe SRS's efforts to reduce the conservatism in the analysis, which resulted in a significantly larger C-14 disposal limit. The work consisted of refining the gas-phase analysis to simulate the migration of C-14 from the waste to the ground surface and evaluated the efficacy of carbonate chemistry in cementitious environment of the ILV for suppressing the volatilization of C-14. During the past year, a Special Analysis was prepared for Department of Energy approval to incorporate the results of these activities that increased the C-14 disposal limits for the ILV, thus allowing for disposal of the Reactor Moderator De-ionizers. Once the Special Analysis is approved by DOE, the actual disposal would be dependent on priority and funding, but the de-ionizers will be removed from the 'waste with no path to disposal list'. (authors)« less

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

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

Mynatt, F.R.

1987-03-18

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

78 FR 13030 - Applications for New Awards; Native American Career and Technical Education Program (NACTEP)

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-26

... DEPARTMENT OF EDUCATION Applications for New Awards; Native American Career and Technical...: Notice. Overview Information: Native American Career and Technical Education Program (NACTEP). Notice... Purpose of Program: The Native American Career and Technical Education Program (NACTEP) provides grants to...

Modifications to the NRAD Reactor, 1977 to present

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

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

1986-01-01

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

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

NASA Technical Reports Server (NTRS)

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

1976-01-01

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

Design of virtual SCADA simulation system for pressurized water reactor

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

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

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

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

DTIC Science & Technology

1993-01-01

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

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

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

Rizwan-uddin; Nick Karancevic; Stefano Markidis

2008-04-23

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

Performance Testing of a Photocatalytic Oxidation Module for Spacecraft Cabin Atmosphere Revitalization

NASA Technical Reports Server (NTRS)

Perry, Jay L.; Abney, Morgan B.; Frederick, Kenneth R.; Scott, Joseph P.; Kaiser, Mark; Seminara, Gary; Bershitsky, Alex

2011-01-01

Photocatalytic oxidation (PCO) is a candidate process technology for use in high volumetric flow rate trace contaminant control applications in sealed environments. The targeted application for PCO as applied to crewed spacecraft life support system architectures is summarized. Technical challenges characteristic of PCO are considered. Performance testing of a breadboard PCO reactor design for mineralizing polar organic compounds in a spacecraft cabin atmosphere is described. Test results are analyzed and compared to results reported in the literature for comparable PCO reactor designs.

75 FR 29732 - Career and Technical Education Program-Promoting Rigorous Career and Technical Education Programs...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-27

... DEPARTMENT OF EDUCATION Career and Technical Education Program--Promoting Rigorous Career and Technical Education Programs of Study Catalog of Federal Domestic Assistance (CFDA) Number: 84.051C. AGENCY: Office of Vocational and Adult Education, Department of Education. ACTION: Notice of proposed priorities...

Liquid Metal Pump Technologies for Nuclear Surface Power

NASA Technical Reports Server (NTRS)

Polzin, Kurt A.

2007-01-01

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

Baseline Concept Description of a Small Modular High Temperature Reactor

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

Hans Gougar

2014-05-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNPmore » were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.« less

Baseline Concept Description of a Small Modular High Temperature Reactor

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

Gougar, Hans D.

2014-10-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNPmore » were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.« less

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

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

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

1991-01-01

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

A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

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

Jacques V Hugo; David I Gertman; Jeffrey C Joe

2014-08-01

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

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

NASA Astrophysics Data System (ADS)

Palmiste, Ü.; Voll, H.

2017-10-01

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

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

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

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

2003-02-25

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

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

Nuclear Engine System Simulation (NESS) version 2.0

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-01-01

The topics are presented in viewgraph form and include the following; nuclear thermal propulsion (NTP) engine system analysis program development; nuclear thermal propulsion engine analysis capability requirements; team resources used to support NESS development; expanded liquid engine simulations (ELES) computer model; ELES verification examples; NESS program development evolution; past NTP ELES analysis code modifications and verifications; general NTP engine system features modeled by NESS; representative NTP expander, gas generator, and bleed engine system cycles modeled by NESS; NESS program overview; NESS program flow logic; enabler (NERVA type) nuclear thermal rocket engine; prismatic fuel elements and supports; reactor fuel and support element parameters; reactor parameters as a function of thrust level; internal shield sizing; and reactor thermal model.

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

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

Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang

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

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

NASA Astrophysics Data System (ADS)

Poston, David I.

2003-01-01

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

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

Light Water Reactor Sustainability Program: Integrated Program Plan

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

None, None

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

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

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

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

Function of university reactors in operator licensing training for nuclear utilities

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

Wicks, F.

1985-11-01

The director of the Division of the US Nuclear Regulatory Commission in generic letter 84-10, dated April 26, 1984, spoke the requirement that applicants for senior reactor operator licenses for power reactors shall have performed then reactor startups. Simulator startups were not acknowledged. Startups performed on a university reactor are acceptable. The content and results of a five-day program combining instruction and experiments with the Rensselaer reactor are summarized.

Down-selection of candidate alloys for further testing of advanced replacement materials for LWR core internals

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

Was, Gary; Leonard, Keith J.; Tan, Lizhen

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) Light Water Reactor Sustainability Program to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to identify and develop advanced alloys with superiormore » degradation resistance in light water reactor (LWR)-relevant environments by 2024.« less

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

Stillman, J. A.; Feldman, E. E.; Wilson, E. H.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains themore » results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo (U-10Mo).« less

Fusion materials: Technical evaluation of the technology of vandium alloys for use as blanket structural materials in fusion power systems

NASA Astrophysics Data System (ADS)

1993-08-01

The Committee's evaluation of vanadium alloys as a structural material for fusion reactors was constrained by limited data and time. The design of the International Thermonuclear Experimental Reactor is still in the concept stage, so meaningful design requirements were not available. The data on the effect of environment and irradiation on vanadium alloys were sparse, and interpolation of these data were made to select the V-5Cr-5Ti alloy. With an aggressive, fully funded program it is possible to qualify a vanadium alloy as the principal structural material for the ITER blanket in the available 5 to 8-year window. However, the data base for V-5Cr-5Ti is limited and will require an extensive development and test program. Because of the chemical reactivity of vanadium the alloy will be less tolerant of system failures, accidents, and off-normal events than most other candidate blanket structural materials and will require more careful handling during fabrication of hardware. Because of the cost of the material more stringent requirements on processes, and minimal historical working experience, it will cost an order of magnitude to qualify a vanadium alloy for ITER blanket structures than other candidate materials. The use of vanadium is difficult and uncertain; therefore, other options should be explored more thoroughly before a final selection of vanadium is confirmed. The Committee views the risk as being too high to rely solely on vanadium alloys. In viewing the state and nature of the design of the ITER blanket as presented to the Committee, it is obvious that there is a need to move toward integrating fabrication, welding, and materials engineers into the ITER design team. If the vanadium alloy option is to be pursued, a large program needs to be started immediately. The commitment of funding and other resources needs to be firm and consistent with a realistic program plan.

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

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

Wolff, Dietmar; Voelzke, Holger; Weber, Wolfgang

2007-07-01

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

Development of NASA Technical Standards Program Relative to Enhancing Engineering Capabilities

NASA Technical Reports Server (NTRS)

Gill, Paul S.; Vaughan, William W.

2003-01-01

The enhancement of engineering capabilities is an important aspect of any organization; especially those engaged in aerospace development activities. Technical Standards are one of the key elements of this endeavor. The NASA Technical Standards Program was formed in 1997 in response to the NASA Administrator s directive to develop an Agencywide Technical Standards Program. The Program s principal objective involved the converting Center-unique technical standards into Agency wide standards and the adoption/endorsement of non-Government technical standards in lieu of government standards. In the process of these actions, the potential for further enhancement of the Agency s engineering capabilities was noted relative to value of being able to access Agencywide the necessary full-text technical standards, standards update notifications, and integration of lessons learned with technical standards, all available to the user from one Website. This was accomplished and is now being enhanced based on feedbacks from the Agency's engineering staff and supporting contractors. This paper addresses the development experiences with the NASA Technical Standards Program and the enhancement of the Agency's engineering capabilities provided by the Program s products. Metrics are provided on significant aspects of the Program.

Alternate Tritium Production Methods Using A Liquid Lithium Target

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

Wilson, J.

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

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

NASA Astrophysics Data System (ADS)

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

2000-12-01

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

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

NASA Technical Reports Server (NTRS)

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

2005-01-01

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

REACTOR PHYSICS CONSTANTS

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

None

1963-07-01

This second edition is based on data available on March 15, 1961. Sections on constants necessary for the interpretation of experimental data and on digital computer programs for reactor design and reactor physics have been added. 1344 references. (D.C.W.)

Policies and practices pertaining to the selection, qualification requirements, and training programs for nuclear-reactor operating personnel at the Oak Ridge National Laboratory

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

Culbert, W.H.

1985-10-01

This document describes the policies and practices of the Oak Ridge National Laboratory (ORNL) regarding the selection of and training requirements for reactor operating personnel at the Laboratory's nuclear-reactor facilities. The training programs, both for initial certification and for requalification, are described and provide the guidelines for ensuring that ORNL's research reactors are operated in a safe and reliable manner by qualified personnel. This document gives an overview of the reactor facilities and addresses the various qualifications, training, testing, and requalification requirements stipulated in DOE Order 5480.1A, Chapter VI (Safety of DOE-Owned Reactors); it is intended to be in compliancemore » with this DOE Order, as applicable to ORNL facilities. Included also are examples of the documentation maintained amenable for audit.« less

Neutrino Physics with Nuclear Reactors: An Overview

NASA Astrophysics Data System (ADS)

Ochoa-Ricoux, J. P.

Nuclear reactors provide an excellent environment for studying neutrinos and continue to play a critical role in unveiling the secrets of these elusive particles. A rich experimental program with reactor antineutrinos is currently ongoing, and leads the way in precision measurements of several oscillation parameters and in searching for new physics, such as the existence of light sterile neutrinos. Ongoing experiments have also been able to measure the flux and spectral shape of reactor antineutrinos with unprecedented statistics and as a function of core fuel evolution, uncovering anomalies that will lead to new physics and/or to an improved understanding of antineutrino emission from nuclear reactors. The future looks bright, with an aggressive program of next generation reactor neutrino experiments that will go after some of the biggest open questions in the field. This includes the JUNO experiment, the largest liquid scintillator detector ever constructed which will push the limits of this detection technology.

R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

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

Vanderhaegen, M.; Laboratory of Waves and Acoustic, Institut Langevin, ESPCI ParisTech, 10 rue Vauquelin, 75005 Paris; Paumel, K.

2011-07-01

In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boilingmore » intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)« less

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

Mulder, R.U.; Benneche, P.E.; Hosticka, B.

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA from use at their institutions. These areas are discussed in this report.« less

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

NONE

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed further in the report.« less

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

Mulder, R.U.; Benneche, P.E.; Hosticka, B.

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here.« less

10 CFR 455.122 - Applicant certifications for State grants for technical assistance, program assistance, and...

Code of Federal Regulations, 2011 CFR

2011-01-01

... assistance, program assistance, and marketing. 455.122 Section 455.122 Energy DEPARTMENT OF ENERGY ENERGY... for State grants for technical assistance, program assistance, and marketing. Applications from States for financial assistance for technical assistance programs, program assistance, and marketing shall...

Multi-University Southeast INIE Consortium

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

Ayman Hawari; Nolan Hertel; Mohamed Al-Sheikhly

2 Project Summary: The Multi-University Southeast INIE Consortium (MUSIC) was established in response to the US Department of Energy’s (DOE) Innovations in Nuclear Infrastructure and Education (INIE) program. MUSIC was established as a consortium composed of academic members and national laboratory partners. The members of MUSIC are the nuclear engineering programs and research reactors of Georgia Institute of Technology (GIT), North Carolina State University (NCSU), University of Maryland (UMD), University of South Carolina (USC), and University of Tennessee (UTK). The University of Florida (UF), and South Carolina State University (SCSU) were added to the MUSIC membership in the second year.more » In addition, to ensure proper coordination between the academic community and the nation’s premier research and development centers in the fields of nuclear science and engineering, MUSIC created strategic partnerships with Oak Ridge National Laboratory (ORNL) including the Spallation Neutron Source (SNS) project and the Joint Institute for Neutron Scattering (JINS), and the National Institute of Standards and Technology (NIST). A partnership was also created with the Armed Forces Radiobiology Research Institute (AFRRI) with the aim of utilizing their reactor in research if funding becomes available. Consequently, there are three university research reactors (URRs) within MUSIC, which are located at NCSU (1-MW PULSTAR), UMD (0.25-MW TRIGA) and UF (0.10-MW Argonaut), and the AFRRI reactor (1-MW TRIGA MARK F). The overall objectives of MUSIC are: a) Demonstrate that University Research Reactors (URR) can be used as modern and innovative instruments of research in the basic and applied sciences, which include applications in fundamental physics, materials science and engineering, nondestructive examination, elemental analysis, and contributions to research in the health and medical sciences, b) Establish a strong technical collaboration between the nuclear engineering faculty and the MUSIC URRs. This will be achieved by involving the faculty in the development of state-of-the-art research facilities at the URRs and subsequently, in the utilization of these facilities, c) Facilitate the use of the URRs by the science and engineering faculty within the individual institutions and by the general community of science and engineering, d) Develop a far-reaching educational component that is capable of addressing the needs of the nuclear science and engineering community. Specifically, the aim of this component will be to perform public outreach activities, contribute to the active recruitment of the next generation of nuclear professionals, strengthen the education of nuclear engineering students, and promote nuclear engineering education for minority students.« less

13 CFR 124.701 - What is the purpose of the 7(j) management and technical assistance program?

Code of Federal Regulations, 2010 CFR

2010-01-01

...) management and technical assistance program? 124.701 Section 124.701 Business Credit and Assistance SMALL...) Business Development Management and Technical Assistance Program § 124.701 What is the purpose of the 7(j) management and technical assistance program? Section 7(j)(1) of the Small Business Act, 15 U.S.C. 636(j)(1...

Technical Skill Attainment and Post-Program Outcomes: An Analysis of Pennsylvania Secondary Career and Technical Education Graduates

ERIC Educational Resources Information Center

Staklis, Sandra; Klein, Steven

2010-01-01

Since the mid-1990s, the Pennsylvania Department of Education (PDE) has required all students concentrating in career and technical education (CTE) programs to complete a standardized technical skill assessment at or near the end of their program. Results of technical skill assessments are used for a number of purposes, including recognizing…

Proton exchange membrane fuel cell technology for transportation applications

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

Swathirajan, S.

1996-04-01

Proton Exchange Membrane (PEM) fuel cells are extremely promising as future power plants in the transportation sector to achieve an increase in energy efficiency and eliminate environmental pollution due to vehicles. GM is currently involved in a multiphase program with the US Department of Energy for developing a proof-of-concept hybrid vehicle based on a PEM fuel cell power plant and a methanol fuel processor. Other participants in the program are Los Alamos National Labs, Dow Chemical Co., Ballard Power Systems and DuPont Co., In the just completed phase 1 of the program, a 10 kW PEM fuel cell power plantmore » was built and tested to demonstrate the feasibility of integrating a methanol fuel processor with a PEM fuel cell stack. However, the fuel cell power plant must overcome stiff technical and economic challenges before it can be commercialized for light duty vehicle applications. Progress achieved in phase I on the use of monolithic catalyst reactors in the fuel processor, managing CO impurity in the fuel cell stack, low-cost electrode-membrane assembles, and on the integration of the fuel processor with a Ballard PEM fuel cell stack will be presented.« less

Materials technology for an advanced space power nuclear reactor concept: Program summary

NASA Technical Reports Server (NTRS)

Gluyas, R. E.; Watson, G. K.

1975-01-01

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

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

NASA Technical Reports Server (NTRS)

Weinstein, H.; Lavan, Z.

1975-01-01

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

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

Development of advanced strain diagnostic techniques for reactor environments.

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

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

2013-02-01

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

Space-based power conversion and power relay systems: Preliminary analysis of alternate systems

NASA Technical Reports Server (NTRS)

1976-01-01

The results are presented of nine months of technical study of non-photovoltaic options for the generation of electricity for terrestrial use by satellite power stations (SPS). A concept for the augmentation of ground-based solar power plants by orbital sunlight reflectors was also studied. Three SPS types having a solar energy source and two which used nuclear reactors were investigated. Data derived for each included: (1) configuration definition, including mass statement; (2) information for use in environmental impact assessment; (3) energy balance (ratio of energy produced to that required to achieve operation), and (4) development and other cost estimates. Cost estimates were dependent upon the total program (development, placement and operation of a number of satellites) which was postulated. This postulation was based upon an analysis of national power capacity trends and guidelines received from MSFC.

ECUT (Energy Conversion and Utilization Technologies) program: Biocatalysis Project

NASA Technical Reports Server (NTRS)

1988-01-01

Fiscal year 1987 research activities and accomplishments for the Biocatalysis Project of the U.S. Department of Energy, Energy Conversion and Utilization Technologies (ECUT) Division are presented. The project's technical activities were organized into three work elements. The Molecular Modeling and Applied Genetics work element includes modeling and simulation studies to verify a dynamic model of the enzyme carboxypeptidase; plasmid stabilization by chromosomal integration; growth and stability characteristics of plasmid-containing cells; and determination of optional production parameters for hyper-production of polyphenol oxidase. The Bioprocess Engineering work element supports efforts in novel bioreactor concepts that are likely to lead to substantially higher levels of reactor productivity, product yields, and lower separation energetics. The Bioprocess Design and Assessment work element attempts to develop procedures (via user-friendly computer software) for assessing the economics and energetics of a given biocatalyst process.

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

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

Bayham, Sanuel; Straub, Doug; Weber, Justin

2017-02-01

As part of the U.S. Department of Energy’s Advanced Combustion Program, the National Energy Technology Laboratory’s Research and Innovation Center (NETL R&IC) is investigating the feasibility of a novel combustion concept in which the GHG emissions can be significantly reduced. This concept involves burning fuel and air without mixing these two reactants. If this concept is technically feasible, then CO 2 emissions can be significantly reduced at a much lower cost than more conventional approaches. This indirect combustion concept has been called Chemical Looping Combustion (CLC) because an intermediate material (i.e., a metaloxide) is continuously cycled to oxidize the fuel.more » This CLC concept is the focus of this research and will be described in more detail in the following sections.« less

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

NASA Technical Reports Server (NTRS)

Dominguez, Jesus; Sibille, Laurent

2010-01-01

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

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

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

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

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

DOE standard: The Department of Energy Laboratory Accreditation Program for radiobioassay

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

NONE

1998-12-01

This technical standard describes the US Department of Energy Laboratory Accreditation Program (DOELAP) for Radiobioassay, for use by the US Department of Energy (DOE) and DOE Contractor radiobioassay programs. This standard is intended to be used in conjunction with the general administrative technical standard that describes the overall DOELAP accreditation process--DOE-STD-1111-98, Department of Energy Laboratory Accreditation Program Administration. This technical standard pertains to radiobioassay service laboratories that provide either direct or indirect (in vivo or in vitro) radiobioassay measurements in support of internal dosimetry programs at DOE facilities or for DOE and DOE contractors. Similar technical standards have been developedmore » for other DOELAP dosimetry programs. This program consists of providing an accreditation to DOE radiobioassay programs based on successful completion of a performance-testing process and an on-site evaluation by technical experts. This standard describes the technical requirements and processes specific to the DOELAP Radiobioassay Accreditation Program as required by 10 CFR 835 and as specified generically in DOE-STD-1111-98.« less

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

PubMed

di Norcia, Vincent

2002-01-01

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

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

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

Cottrell, W.B.; Passiakos, M.

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

A Neutronic Program for Critical and Nonequilibrium Study of Mobile Fuel Reactors: The Cinsf1D Code

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

Lecarpentier, David; Carpentier, Vincent

2003-01-15

Molten salt reactors (MSRs) have the distinction of having a liquid fuel that is also the coolant. The transport of delayed-neutron precursors by the fuel modifies the precursors' equation. As a consequence, it is necessary to adapt the methods currently used for solid fuel reactors to achieve critical or kinetics calculations for an MSR. A program is presented for which this adaptation has been carried out within the framework of the two-energy-group diffusion theory with one dimension of space. This program has been called Cinsf1D (Cinetique pour reacteur a sels fondus 1D)

76 FR 35474 - UAW-Chrysler Technical Training Center, Technology Training Joint Programs Staff, Including On...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-17

...-Chrysler Technical Training Center, Technology Training Joint Programs Staff, Including On-Site Leased Workers From Cranks, O/E Learning, DBSI, IDEA, and Tonic/MVP, Detroit, MI; UAW-Chrysler Technical Training... workers and former workers of UAW-Chrysler Technical Training Center, Technology Training Joint Programs...

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-31

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

Determination of the Arrhenius Activation Energy Using a Temperature-Programmed Flow Reactor.

ERIC Educational Resources Information Center

Chan, Kit-ha C.; Tse, R. S.

1984-01-01

Describes a novel method for the determination of the Arrhenius activation energy, without prejudging the validity of the Arrhenius equation or the concept of activation energy. The method involves use of a temperature-programed flow reactor connected to a concentration detector. (JN)

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

Chemical Technology Division annual technical report, 1990

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

Not Available

1991-05-01

Highlights of the Chemical Technology (CMT) Division's activities during 1990 are presented. In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for coal- fired magnetohydrodynamics and fluidized-bed combustion; (3) methods for recovery of energy from municipal waste and techniques for treatment of hazardous organic waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for a high-level waste repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams, concentrating plutonium solids in pyrochemical residues by aqueous biphase extraction, andmore » treating natural and process waters contaminated by volatile organic compounds; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (IFR); (7) processes for removal of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials in environments simulating those of fission and fusion energy systems. The Division also has a program in basic chemistry research in the areas of fluid catalysis for converting small molecules to desired products; materials chemistry for superconducting oxides and associated and ordered solutions at high temperatures; interfacial processes of importance to corrosion science, high-temperature superconductivity, and catalysis; and the geochemical processes responsible for trace-element migration within the earth's crust. The Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the scientific and engineering programs at Argonne National Laboratory (ANL). 66 refs., 69 figs., 6 tabs.« less

Nuclear Technology Series. Nuclear Reactor (Plant) Operator Trainee. A Suggested Program Planning Guide. Revised June 80.

ERIC Educational Resources Information Center

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

This program planning guide for a two-year postsecondary nuclear reactor (plant) operator trainee program is designed for use with courses 1-16 of thirty-five in the Nuclear Technology Series. The purpose of the guide is to describe the nuclear power field and its job categories for specialists, technicians and operators; and to assist planners,…

The Secretary's Awards for Outstanding Vocational-Technical Education Programs. 1995.

ERIC Educational Resources Information Center

Office of Vocational and Adult Education (ED), Washington, DC.

This publication describes the 12 winning programs and the 9 semifinalists of the 1995 U.S. Secretary of Education's Awards for Outstanding Vocational-Technical Education Programs. The awards program was established in 1981 to recognize excellence in local school-to-work and vocational-technical education programs, services, and activities that…

75 FR 29366 - ``Homeless Veterans' Reintegration Program (HVRP) National Technical Assistance Center...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-25

... ``Homeless Veterans' Reintegration Program (HVRP) National Technical Assistance Center Cooperative Agreement...). Section 2021 authorizes programs to expedite the reintegration of homeless Veterans into the labor force... Technical Assistance Center (NTAC) for the Homeless Veterans' Reintegration Program (HVRP) to include the...

KiloPower Program

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

McClure, Patrick Ray

2016-08-04

These are the slides for a phone interview with Aerospace America magazine of the AIAA. It goes over the KiloPower Program at Los Alamos National Laboratory (LANL), and covers the following: 1 kWe Kilopower, 10 kWe Kilopower, Kilopower Reactor Using Stirling Technology (KRUSTY) Integration Test (DAF), Reactor Configuration, and Platen Positions.

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

PubMed

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

2016-12-01

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

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

NASA Astrophysics Data System (ADS)

Rahayu, Suparni Setyowati; Budiyono; Purwanto

2018-02-01

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

78 FR 25292 - Announcement of Funding Awards; Office of Native American Programs Training and Technical...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-30

... Awards; Office of Native American Programs Training and Technical Assistance; Fiscal Year 2012 AGENCY... (NOFA) for the Office of Native American Programs Training and Technical Assistance (ONAP T&TA). This... nonprofit organizations, as well as for-profit entities to provide Training & Technical Assistance to the...

Low Cost Technical Solutions to Jump Start an Insider Threat Program

DTIC Science & Technology

2016-05-11

Low Cost Technical Solutions to Jump Start an Insider Threat Program George J. Silowash Derrick L. Spooner Daniel L. Costa Michael J...Albrethsen May 2016 TECHNICAL NOTE CMU/SEI-2016-TN-004 CERT Division http://www.sei.cmu.edu Copyright 2016 Carnegie Mellon University This... technical note will explore tools that may be suitable for satisfying the basic technical needs of an insider threat program, giving organizations a

Small reactor power systems for manned planetary surface bases

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.

1987-01-01

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

VVER Reactor Safety in Eastern Europe and Former Soviet Union

NASA Astrophysics Data System (ADS)

Papadopoulou, Demetra

2012-02-01

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

Owning the program technical baseline for future space systems acquisition: program technical baseline tracking tool

NASA Astrophysics Data System (ADS)

Nguyen, Tien M.; Guillen, Andy T.; Hant, James J.; Kizer, Justin R.; Min, Inki A.; Siedlak, Dennis J. L.; Yoh, James

2017-05-01

The U.S. Air Force (USAF) has recognized the needs for owning the program and technical knowledge within the Air Force concerning the systems being acquired to ensure success. This paper extends the previous work done by the authors [1-2] on the "Resilient Program Technical Baseline Framework for Future Space Systems" and "Portfolio Decision Support Tool (PDST)" to the development and implementation of the Program and Technical Baseline (PTB) Tracking Tool (PTBTL) for the DOD acquisition life cycle. The paper describes the "simplified" PTB tracking model with a focus on the preaward phases and discusses how to implement this model in PDST.

NASA Technical Standards Program

NASA Technical Reports Server (NTRS)

Gill, Paul S.; Vaughan, William W.; Parker, Nelson C. (Technical Monitor)

2002-01-01

The NASA Technical Standards Program was officially established in 1997 as result of a directive issued by the Administrator. It is responsible for Agency wide technical standards development, adoption (endorsement), and conversion of Center-unique standards for Agency wide use. One major element of the Program is the review of NSA technical standards products and replacement with non-Government Voluntary Consensus Standards in accordance with directions issued by the Office of Management and Budget. As part of the Program's function, it developed a NASA Integrated Technical Standards Initiative that consists of and Agency wide full-text system, standards update notification system, and lessons learned-standards integration system. The Program maintains a 'one stop-shop' Website for technical standards ad related information on aerospace materials, etc. This paper provides information on the development, current status, and plans for the NAS Technical Standards Program along with metrics on the utility of the products provided to both users within the nasa.gov Domain and the Public Domain.

NASA Technical Standards Program

NASA Technical Reports Server (NTRS)

Gill, Paul S.; Vaughan, WIlliam W.

2003-01-01

The NASA Technical Standards Program was officially established in 1997 as result of a directive issued by the Administrator. It is responsible for Agency wide technical standards development, adoption (endorsement), and conversion of Center-unique standards for Agency wide use. One major element of the Program is the review of NSA technical standards products and replacement with non-Government Voluntary Consensus Standards in accordance with directions issued by the Office of Management and Budget. As part of the Program s function, it developed a NASA Integrated Technical Standards Initiative that consists of and Agency wide full-text system, standards update notification system, and lessons learned - standards integration system. The Program maintains a "one stop-shop" Website for technical standards ad related information on aerospace materials, etc. This paper provides information on the development, current status, and plans for the NAS Technical Standards Program along with metrics on the utility of the products provided to both users within the nasa.gov Domain and the Public Domain.

Middlesex Community College Software Technical Writing Program.

ERIC Educational Resources Information Center

Middlesex Community Coll., Bedford, MA.

This document describes the Software Technical Writing Program at Middlesex Community College (Massachusetts). The program is a "hands-on" course designed to develop job-related skills in three major areas: technical writing, software, and professional skills. The program was originally designed in cooperation with the Massachusetts High…

Technical Assistance and Changes in Nutrition and Physical Activity Practices in the National Early Care and Education Learning Collaboratives Project, 2015-2016.

PubMed

Chiappone, Alethea; Smith, Teresa M; Estabrooks, Paul A; Rasmussen, Cristy Geno; Blaser, Casey; Yaroch, Amy L

2018-04-26

The National Early Care and Education Learning Collaboratives Project (ECELC) aims to improve best practices in early care and education (ECE) programs in topic areas of the Nutrition and Physical Activity Self-Assessment in Child Care (NAP SACC). Technical assistance is a component of the ECELC, yet its effect on outcomes is unclear. Beyond dose and duration of technical assistance, limited research exists on characteristics of technical assistance that contribute to outcomes. The objective of this study was to identify and describe technical assistance characteristics and explore associations with NAP SACC outcomes. We collected data from 10 collaboratives comprising 84 ECE programs in 2 states in 2015-2016. The objective of technical assistance was to support programs in improving best practices. Technical assistance was provided to programs via on-site, telephone, or email and was tailored to program needs. We used a mixed-methods design to examine associations between technical assistance and NAP SACC outcomes. We used multiple regression analysis to assess quantitative data and qualitative comparative analysis to determine necessary and sufficient technical assistance conditions supporting NAP SACC outcomes. We also conducted a document review to describe technical assistance that referred conditions identified by the qualitative comparative analysis. Regression analyses detected an inverse relationship between changes in NAP SACC scores and hours of technical assistance. No clear pattern emerged in the qualitative comparative analysis, leaving no necessary and sufficient conditions. However, the qualitative comparative analysis identified feedback as a potentially important component of technical assistance, whereas resource sharing and frequent email were characteristics that seemed to reduce the likelihood of improved outcomes. Email and resource sharing were considered primarily general information rather than tailored technical assistance. Technical assistance may be used in programs and made adaptable to program needs. The inclusion and evaluation of technical assistance, especially tailored approaches, is warranted for environmental interventions, including ECE settings.

Development of a three-dimensional core dynamics analysis program for commercial boiling water reactors

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

Bessho, Yasunori; Yokomizo, Osamu; Yoshimoto, Yuichiro

1997-03-01

Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates calculational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory size as attained by the latest supercomputer technology. The program consists of six major modules, namely a core neutronics module, a fuel heat conduction/transfer module, a fuel channel thermal-hydraulic module, an upper plenum/separator module, a feedwater/recirculation flow module, and amore » control system module. Its core neutronics module is based on the modified one-group neutron kinetics equation with the prompt jump approximation and with six delayed neutron precursor groups. The module is used to analyze one fuel bundle of the reactor core with one mesh (region). The fuel heat conduction/transfer module solves the one-dimensional heat conduction equation in the radial direction with ten nodes in the fuel pin. The fuel channel thermal-hydraulic module is based on separated three-equation, two-phase flow equations with the drift flux correlation, and it analyzes one fuel bundle of the reactor core with one channel to evaluate flow redistribution between channels precisely. Thermal margin is evaluated by using the GEXL correlation, for example, in the module.« less

Report Summarizing the Effort Required to Initiate Welding of Irradiated Materials within the Welding Cubicle

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

Frederick, Greg; Sutton, Benjamin J.; Tatman, Jonathan K.

The advanced welding facility within a hot cell at the Radiochemical Engineering Development Center of Oak Ridge National Laboratory (ORNL), which has been jointly funded by the U.S. Department of Energy (DOE), Office of Nuclear Energy, Light Water Reactor Sustainability Program and the Electric Power Research Institute, Long Term Operations Program and the Welding and Repair Technology Center, is in the final phase of development. Research and development activities in this facility will involve direct testing of advanced welding technologies on irradiated materials in order to address the primary technical challenge of helium induced cracking that can arise when conventionalmore » fusion welding techniques are utilized on neutron irradiated stainless steels and nickel-base alloys. This report details the effort that has been required since the beginning of fiscal year 2017 to initiate welding research and development activities on irradiated materials within the hot cell cubicle, which houses welding sub-systems that include laser beam welding (LBW) and friction stir welding (FSW) and provides material containment within the hot cell.« less

Automated Work Packages Prototype: Initial Design, Development, and Evaluation. Light Water Reactor Sustainability Program

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

Oxstrand, Johanna Helene; Ahmad Al Rashdan; Le Blanc, Katya Lee

The goal of the Automated Work Packages (AWP) project is to demonstrate how to enhance work quality, cost management, and nuclear safety through the use of advanced technology. The work described in this report is part of the digital architecture for a highly automated plant project of the technical program plan for advanced instrumentation, information, and control (II&C) systems technologies. This report addresses the DOE Milestone M2LW-15IN0603112: Describe the outcomes of field evaluations/demonstrations of the AWP prototype system and plant surveillance and communication framework requirements at host utilities. A brief background to the need for AWP research is provided, thenmore » two human factors field evaluation studies are described. These studies focus on the user experience of conducting a task (in this case a preventive maintenance and a surveillance test) while using an AWP system. The remaining part of the report describes an II&C effort to provide real time status updates to the technician by wireless transfer of equipment indications and a dynamic user interface.« less

A Physicist's Journey In The Nuclear Power World

NASA Astrophysics Data System (ADS)

Starr, Chauncey

2000-03-01

As a participant in the development of civilian nuclear power plants for the past half century, the author presents some of his insights to its history that may be of interest to today's applied physicists. Nuclear power development has involved a mixture of creative vision, science, engineering, and unusual technical, economic, and social obstacles. Nuclear power programs were initiated during the euphoric era of public support for new science immediately following World War II -- a support that lasted almost two decades. Subsequently, nuclear power has had to face a complex mix of public concerns and criticism. The author's involvment in some of these circumstances will be anecdotally described. Although the physics of fission and its byproducts remains at the heart of all nuclear reactor designs, its embodiment in practical energy sources has been shaped by the limitations of engineering primarily and economics secondarily. Very influential has been the continuing interplay with the military's weapons and propulsion programs, and the government's political policies. In this respect, nuclear power's history provides a learning experience that may be applicable to some of the large scale demonstration projects that physicists pursue today.

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

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

Nelson, George W.

1986-07-01

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

Guidelines for Establishing and Evaluating High School Technical Electromechanics Programs.

ERIC Educational Resources Information Center

Florida State Dept. of Education, Tallahassee. Div. of Vocational, Technical and Adult Education.

Educators and industrial representatives developed these guidelines for school officials, instructors in technical education, and program and facility planners to use in planning a high school program in technical electromechanics. Designed to train students for entry into industry in applied electromechanics, the program includes electricity,…

Status and problems of fusion reactor development.

PubMed

Schumacher, U

2001-03-01

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

Technical Preparation: Preparing Students for Success. A Position Paper Presented by the Wisconsin Association of Secondary Vocational Administrators 1991-1992.

ERIC Educational Resources Information Center

Peterson, Charlene; And Others

Technical Preparation (Tech Prep) programs include basic skills and competencies to improve students' school-to-work transition. The following elements characterize Tech Prep: designed for grades 9-14, leading to an technical associate degree; parallels college preparatory program; offers technical program of study in business enterprise,…

Leveraging Independent Management and Chief Engineer Hierarchy: Vertically and Horizontally-Derived Technical Authority Value

NASA Technical Reports Server (NTRS)

Barley, Bryan; Newhouse, Marilyn

2012-01-01

In the development of complex spacecraft missions, project management authority is usually extended hierarchically from NASA's highest agency levels down to the implementing institution's project team level, through both the center and the program. In parallel with management authority, NASA utilizes a complementary, but independent, hierarchy of technical authority (TA) that extends from the agency level to the project, again, through both the center and the program. The chief engineers (CEs) who serve in this technical authority capacity oversee and report on the technical status and ensure sound engineering practices, controls, and management of the projects and programs. At the lowest level, implementing institutions assign project CEs to technically engage projects, lead development teams, and ensure sound technical principles, processes, and issue resolution. At the middle level, programs and centers independently use CEs to ensure the technical success of their projects and programs. At the agency level, NASA's mission directorate CEs maintain technical cognizance over every program and project in their directorate and advise directorate management on the technical, cost, schedule, and programmatic health of each. As part of this vertically-extended CE team, a program level CE manages a continually varying balance between penetration depth and breadth across his or her assigned missions. Teamwork issues and information integration become critical for management at all levels to ensure value-added use of both the synergy available between CEs at the various agency levels, and the independence of the technical authority at each organization.

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

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

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

2012-09-01

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

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

NASA Technical Reports Server (NTRS)

Clark, John S. (Editor)

1991-01-01

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

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

NASA Astrophysics Data System (ADS)

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

2011-04-01

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

Defense Technical Information Center (DTIC) - Its role in the USAF Scientific and Technical Information Program

NASA Technical Reports Server (NTRS)

Kuhn, Allan D.

1991-01-01

The Defense Technical Information Center (DTIC), the central repository for DOD scientific and technical information concerning studies and research and engineering efforts, is discussed. The present makeup of DTIC is described and its functions in producing technical reports and technical report bibliographies are examined. DTIC's outreach services are reviewed, as are its DTIC information and technology transfer programs. DTIC's plans for the year 2000 and its relation to the mission of the U.S. Air Force, including the Air Force's STINFO program, are addressed.

77 FR 37710 - Notice of Proposed Information Collection

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-22

... to continue the collection of information for two technical training program course effectiveness... approval. This collection is for OSM's Technical Training Program Course Effectiveness Evaluations (1029... collection activity: Title: Technical Training Program Course Effectiveness Evaluation. OMB Control Number...

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

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

Michael Tyacke; Frantisek Svitak; Jiri Rychecky

2010-04-01

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

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

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

Michael J. Tyacke; Frantisek Svitak; Jiri Rychecky

2007-10-01

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

Reliability Analysis of RSG-GAS Primary Cooling System to Support Aging Management Program

NASA Astrophysics Data System (ADS)

Deswandri; Subekti, M.; Sunaryo, Geni Rina

2018-02-01

Multipurpose Research Reactor G.A. Siwabessy (RSG-GAS) which has been operating since 1987 is one of the main facilities on supporting research, development and application of nuclear energy programs in BATAN. Until now, the RSG-GAS research reactor has been successfully operated safely and securely. However, because it has been operating for nearly 30 years, the structures, systems and components (SSCs) from the reactor would have started experiencing an aging phase. The process of aging certainly causes a decrease in reliability and safe performances of the reactor, therefore the aging management program is needed to resolve the issues. One of the programs in the aging management is to evaluate the safety and reliability of the system and also screening the critical components to be managed.One method that can be used for such purposes is the Fault Tree Analysis (FTA). In this papers FTA method is used to screening the critical components in the RSG-GAS Primary Cooling System. The evaluation results showed that the primary isolation valves are the basic events which are dominant against the system failure.

Improved Nuclear Reactor and Shield Mass Model for Space Applications

NASA Technical Reports Server (NTRS)

Robb, Kevin

2004-01-01

New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

Comparison of Analysis Results Between 2D/1D Synthesis and RAPTOR-M3G in the Korea Standard Nuclear Plant (KSNP)

NASA Astrophysics Data System (ADS)

Joung Lim, Mi; Maeng, Young Jae; Fero, Arnold H.; Anderson, Stanwood L.

2016-02-01

The 2D/1D synthesis methodology has been used to calculate the fast neutron (E > 1.0 MeV) exposure to the beltline region of the reactor pressure vessel. This method uses the DORT 3.1 discrete ordinates code and the BUGLE-96 cross-section library based on ENDF/B-VI. RAPTOR-M3G (RApid Parallel Transport Of Radiation-Multiple 3D Geometries) which performs full 3D calculations was developed and is based on domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architecture. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor. Both methods are applied to surveillance test results for the Korea Standard Nuclear Plant (KSNP)-OPR (Optimized Power Reactor) 1000 MW. The objective of this paper is to compare the results of the KSNP surveillance program between 2D/1D synthesis and RAPTOR-M3G. Each operating KSNP has a reactor vessel surveillance program consisting of six surveillance capsules located between the core and the reactor vessel in the downcomer region near the reactor vessel wall. In addition to the In-Vessel surveillance program, an Ex-Vessel Neutron Dosimetry (EVND) program has been implemented. In order to estimate surveillance test results, cycle-specific forward transport calculations were performed by 2D/1D synthesis and by RAPTOR-M3G. The ratio between measured and calculated (M/C) reaction rates will be discussed. The current plan is to install an EVND system in all of the Korea PWRs including the new reactor type, APR (Advanced Power Reactor) 1400 MW. This work will play an important role in establishing a KSNP-specific database of surveillance test results and will employ RAPTOR-M3G for surveillance dosimetry location as well as positions in the KSNP reactor vessel.

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

DOT National Transportation Integrated Search

2009-06-01

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

Advanced reactor site feasibility evaluation

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

Haynes, Mark; Owens, Martin; Southworth, Finis

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

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

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

ALLIANCE LIMITED, NGNP INDUSTRY

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

Nuclear Technology Series. Course 7: Reactor Operations.

ERIC Educational Resources Information Center

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

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

Nuclear Technology Series. Course 8: Reactor Safety.

ERIC Educational Resources Information Center

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

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

Nuclear Technology Series. Course 9: Reactor Auxiliary Systems.

ERIC Educational Resources Information Center

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

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

Nuclear Technology Series. Course 12: Reactor Physics.

ERIC Educational Resources Information Center

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

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

Report Card on Nuclear Power

ERIC Educational Resources Information Center

Novick, Sheldon

1974-01-01

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

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

EPA Science Inventory

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

FALCON reactor-pumped laser description and program overview

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

None

1989-12-01

The FALCON (Fission Activated Laser CONcept) reactor-pumped laser program at Sandia National Laboratories is examining the feasibility of high-power systems pumped directly by the energy from a nuclear reactor. In this concept we use the highly energetic fission fragments from neutron induced fission to excite a large volume laser medium. This technology has the potential to scale to extremely large optical power outputs in a primarily self-powered device. A laser system of this type could also be relatively compact and capable of long run times without refueling.

A complete dosimetry experimental program in support to the core characterization and to the power calibration of the CABRI reactor. A complete dosimetry experimental program in support of the core characterization and of the power calibration of the CABRI reactor

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

Rodiac, F.; Hudelot, JP.; Lecerf, J.

CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center. Since 1978 the experimental programs have aimed at studying the fuel behavior under Reactivity Initiated Accident (RIA) conditions. Since 2003, it has been refurbished in order to be able to provide RIA and LOCA (Loss Of Coolant Accident) experiments in prototypical PWR conditions (155 bar, 300 deg. C). This project is part of a broader scope including an overall facility refurbishment and a safety review. The global modification is conducted by the CEA project team. It is funded by IRSN, which is conducting the CIP experimentalmore » program, in the framework of the OECD/NEA project CIP. It is financed in the framework of an international collaboration. During the reactor restart, commissioning tests are realized for all equipment, systems and circuits of the reactor. In particular neutronics and power commissioning tests will be performed respectively in 2015 and 2016. This paper focuses on the design of a complete and original dosimetry program that was built in support to the CABRI core characterization and to the power calibration. Each one of the above experimental goals will be fully described, as well as the target uncertainties and the forecasted experimental techniques and data treatment. (authors)« less

A Special Topic From Nuclear Reactor Dynamics for the Undergraduate Physics Curriculum

ERIC Educational Resources Information Center

Sevenich, R. A.

1977-01-01

Presents an intuitive derivation of the point reactor equations followed by formulation of equations for inverse and direct kinetics which are readily programmed on a digital computer. Suggests several computer simulations involving the effect of control rod motion on reactor power. (MLH)

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

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

Maloy, Stuart Andrew

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

NOAAPORT User's Page

Science.gov Websites

Update History NOAAPORT Technical Contacts NOAAPORT Program Overview NOAAPORT Reference Documentation NOAAPORT Update History - for the NOAAPORT Program. NOAAPORT Technical Contacts - NWS Technical Points of Alignment Information. Receiver Characteristics - Technical Characteristics of the NOAAPORT Receiver. Data

Nuclear Fuel Reprocessing

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

Harold F. McFarlane; Terry Todd

2013-11-01

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

76 FR 18624 - Research, Technical Assistance and Training Programs: Notice of Final Circular

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-04

... to FTA Circular 6100.1D, Research and Technical Assistance Training Program: Application Instructions... DEPARTMENT OF TRANSPORTATION Federal Transit Administration Research, Technical Assistance and Training Programs: Notice of Final Circular AGENCY: Federal Transit Administration (FTA), DOT. ACTION...

10 CFR 455.131 - State ranking of grant applications.

Code of Federal Regulations, 2010 CFR

2010-01-01

...) For technical assistance programs, buildings shall be ranked in descending priority based upon the... all buildings covered by eligible applications for: (1) Technical assistance programs for units of local government and public care institutions and (2) Technical assistance programs for schools and...

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

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

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

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

Coal desulfurization by a microwave process. Technical progress report, February 1981-May 1981

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

Zavitsanos, P.D.; Golden, J.A.; Bleiler, K.W.

1981-01-01

Desulfurization experiments were carried out using the 6KW, 2450 MHz Flow Reactor System. The program has been directed toward the combination of physical separation and microwave exposure with NaOH to increase sulfur removal. The following treatment sequence has been used with good results: (1) expose 1/4 to 1 in. raw coal to microwaves; (2) crush the treated coal and separate the sample into float/sink fractions; (3) add NaOH to the float fraction and re-expose the sample to microwaves; and (4) wash, add NaOH and expose to microwaves. This procedure has produced up to 89% sulfur removal and as low asmore » 0.31 numberS/10/sup 6/ Btu. Ash analyses on these samples showed as high as 40% reduction. The calorific value was increased in almost all samples. Data on sulfur, ash and calorific values are summarized.« less

Sodium-NaK engineering handbook. Volume III. Sodium systems, safety, handling, and instrumentation. [LMFBR

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

Foust, O J

1978-01-01

The handbook is intended for use by present and future designers in the Liquid Metals Fast Breeder Reactor (LMFBR) Program and by the engineering and scientific community performing other type investigation and exprimentation requiring high-temperature sodium and NaK technology. The arrangement of subject matter progresses from a technological discussion of sodium and sodium--potassium alloy (NaK) to discussions of varius categories and uses of hardware in sodium and NaK systems. Emphasis is placed on sodium and NaK as heat-transport media. Sufficient detail is included for basic understanding of sodium and NaK technology and of technical aspects of sodium and NaK componentsmore » and instrument systems. Information presented is considered adequate for use in feasibility studies and conceptual design, sizing components and systems, developing preliminary component and system descriptions, identifying technological limitations and problem areas, and defining basic constraints and parameters.« less

EBR-II and TREAT Digitization Project

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

Griffith, George W.; Rabiti, Cristian

2015-09-01

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

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

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

Percher, Catherine

2012-06-19

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

Liquid-Metal Pump Technologies for Nuclear Surface Power

NASA Technical Reports Server (NTRS)

Polzin, K. A.

2007-01-01

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

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...

Our Next Two Steps for Fukushima Daiichi Muon Tomography

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

Miyadera, Haruo

2012-04-11

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

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

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

Castaing, C.; Cazin, B.

1995-03-01

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

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

NASA Astrophysics Data System (ADS)

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

2016-08-01

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

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

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

Yamada, K.; Aksan, S. N.

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

On-line Naval Engineering Skills Supplemental Training Program

DTIC Science & Technology

2010-01-01

Defense Technical University ( DTU ), the technical content for courses would have to be provided by the Naval technical authorities...of technological knowledge related to design engineering such as the DTU , or expanded within the mission scope of an existing organization such as...management program as a training tool for naval design engineers such as the DTU or a technical extension of the DAU program for acquisition training

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

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

Schenkel, Roland

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

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

NASA Astrophysics Data System (ADS)

Schenkel, Roland

2012-06-01

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

Application of Microprocessor-Based Equipment in Nuclear Power Plants - Technical Basis for a Qualification Methodology

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

Korsah, K.

This document (1) summarizes the most significant findings of the ''Qualification of Advanced Instrumentation and Control (I&C) Systems'' program initiated by the Nuclear Regulatory Commission (NRC); (2) documents a comparative analysis of U.S. and European qualification standards; and (3) provides recommendations for enhancing regulatory guidance for environmental qualification of microprocessor-based safety-related systems. Safety-related I&C system upgrades of present-day nuclear power plants, as well as I&C systems of Advanced Light-Water Reactors (ALWRs), are expected to make increasing use of microprocessor-based technology. The Nuclear Regulatory Commission (NRC) recognized that the use of such technology may pose environmental qualification challenges different from current,more » analog-based I&C systems. Hence, it initiated the ''Qualification of Advanced Instrumentation and Control Systems'' program. The objectives of this confirmatory research project are to (1) identify any unique environmental-stress-related failure modes posed by digital technologies and their potential impact on the safety systems and (2) develop the technical basis for regulatory guidance using these findings. Previous findings from this study have been documented in several technical reports. This final report in the series documents a comparative analysis of two environmental qualification standards--Institute of Electrical and Electronics Engineers (IEEE) Std 323-1983 and International Electrotechnical Commission (IEC) 60780 (1998)--and provides recommendations for environmental qualification of microprocessor-based systems based on this analysis as well as on the findings documented in the previous reports. The two standards were chosen for this analysis because IEEE 323 is the standard used in the U.S. for the qualification of safety-related equipment in nuclear power plants, and IEC 60780 is its European counterpart. In addition, the IEC document was published in 1998, and should reflect any new qualification concerns, from the European perspective, with regard to the use of microprocessor-based safety systems in power plants.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-10

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

77 FR 18271 - Terrestrial Environmental Studies for Nuclear Power Stations

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-27

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

In-service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements

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

Sullivan, Edmund J.; Anderson, Michael T.; Norris, Wallace

2012-09-17

Pressurized thermal shock (PTS) events are system transients in a pressurized water reactor (PWR) in which there is a rapid operating temperature cool-down that results in cold vessel temperatures with or without repressurization of the vessel. The rapid cooling of the inside surface of the reactor pressure vessel (RPV) causes thermal stresses that can combine with stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a pre-existing flaw is present in a material susceptible to brittle failure. The ferritic, low alloy steel of the reactor vessel beltline adjacent tomore » the core, where neutron radiation gradually embrittles the material over the lifetime of the plant, can be susceptible to brittle fracture. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), “Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” adopted on July 23, 1985, establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events” (75 FR 13). Use of the new rule by licensees is optional. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory (PNNL) has been working on a program to assess the ability of current inservice inspection (ISI)-ultrasonic testing (UT) techniques, as qualified through ASME Code, Appendix VIII, Supplements 4 and 6, to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, Spirit of Appendix VIII reactor vessel examination, a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.« less

Technical Assistance and Changes in Nutrition and Physical Activity Practices in the National Early Care and Education Learning Collaboratives Project, 2015–2016

PubMed Central

Smith, Teresa M.; Estabrooks, Paul A.; Rasmussen, Cristy Geno; Blaser, Casey; Yaroch, Amy L

2018-01-01

Purpose and Objectives The National Early Care and Education Learning Collaboratives Project (ECELC) aims to improve best practices in early care and education (ECE) programs in topic areas of the Nutrition and Physical Activity Self-Assessment in Child Care (NAP SACC). Technical assistance is a component of the ECELC, yet its effect on outcomes is unclear. Beyond dose and duration of technical assistance, limited research exists on characteristics of technical assistance that contribute to outcomes. The objective of this study was to identify and describe technical assistance characteristics and explore associations with NAP SACC outcomes. Intervention Approach We collected data from 10 collaboratives comprising 84 ECE programs in 2 states in 2015–2016. The objective of technical assistance was to support programs in improving best practices. Technical assistance was provided to programs via on-site, telephone, or email and was tailored to program needs. Evaluation Methods We used a mixed-methods design to examine associations between technical assistance and NAP SACC outcomes. We used multiple regression analysis to assess quantitative data and qualitative comparative analysis to determine necessary and sufficient technical assistance conditions supporting NAP SACC outcomes. We also conducted a document review to describe technical assistance that referred conditions identified by the qualitative comparative analysis. Results Regression analyses detected an inverse relationship between changes in NAP SACC scores and hours of technical assistance. No clear pattern emerged in the qualitative comparative analysis, leaving no necessary and sufficient conditions. However, the qualitative comparative analysis identified feedback as a potentially important component of technical assistance, whereas resource sharing and frequent email were characteristics that seemed to reduce the likelihood of improved outcomes. Email and resource sharing were considered primarily general information rather than tailored technical assistance. Implications for Public Health Technical assistance may be used in programs and made adaptable to program needs. The inclusion and evaluation of technical assistance, especially tailored approaches, is warranted for environmental interventions, including ECE settings. PMID:29704371

Status of JUPITER Program

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

Inoue, T.; Shirakata, K.; Kinjo, K.

To obtain the data necessary for evaluating the nuclear design method of a large-scale fast breeder reactor, criticality tests with a large- scale homogeneous reactor were conducted as part of a joint research program by Japan and the U.S. Analyses of the tests are underway in both countries. The purpose of this paper is to describe the status of this project.

Technical assistance for hazardous-waste reduction

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

Thompson, F.M.; McComas, C.A.

1987-12-01

Minnesota's Waste Management Board has established, developed, and funded the Minnesota Technical Assistance Program (MnTAP). The MnTAP programs offers technical assistance to generators of hazardous waste by offering telephone and onsite consultation, a waste reduction resource bank, information dissemination, a student intern program, and research awards for waste reduction projects. The program has completed three years of successful operation. The increasing interest in and use of MnTAP's services by hazardous-waste generators has justified the belief that state technical assistance programs have an important role to play in helping generators to reduce their waste production.

15 CFR 296.32 - Technical and educational services.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 15 Commerce and Foreign Trade 1 2010-01-01 2010-01-01 false Technical and educational services... NATIONAL INSTITUTE OF STANDARDS AND TECHNOLOGY, DEPARTMENT OF COMMERCE NIST EXTRAMURAL PROGRAMS TECHNOLOGY INNOVATION PROGRAM Dissemination of Program Results § 296.32 Technical and educational services. (a) Under...

40 CFR 52.1110 - Small business stationary source technical and environmental compliance assistance program.

Code of Federal Regulations, 2010 CFR

2010-07-01

... technical and environmental compliance assistance program. 52.1110 Section 52.1110 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) APPROVAL AND PROMULGATION OF IMPLEMENTATION PLANS (CONTINUED) Maryland § 52.1110 Small business stationary source technical and environmental...

76 FR 14372 - New Mexico Collaborative Forest Restoration Program Technical Advisory Panel

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-16

... DEPARTMENT OF AGRICULTURE Forest Service New Mexico Collaborative Forest Restoration Program Technical Advisory Panel AGENCY: Forest Service, USDA. ACTION: Notice of meeting. SUMMARY: The New Mexico Collaborative Forest Restoration Program Technical Advisory Panel will meet in Albuquerque, New Mexico. The...

How does intensification influence the operational and environmental performance of photo-Fenton processes at acidic and circumneutral pH.

PubMed

Salazar, Luis Miguel; Grisales, Claudia Mildred; Garcia, Dorian Prato

2018-05-31

This study evaluates the technical, economical, and environmental impact of sodium persulfate (Na 2 S 2 O 8 ) as an enhancing agent in a photo-Fenton process within a solar-pond type reactor (SPR). Photo-Fenton (PF) and photo-Fenton intensified with the addition of persulfate (PFPS) processes decolorize 97% the azo dye direct blue 71 (DB71) and allow producing a highly biodegradable effluent. Intensification with persulfate allowed reducing treatment time in 33% (from 120 to 80 min) and the consumption of chemical auxiliaries needed for pH adjustment. Energy, reagents, and chemical auxiliaries are still and environmental hotspot for PF and PFPS; however, it is worth mentioning that their environmental footprint is lower than that observed for compound parabolic concentrator (CPC)-type reactors. A life-cycle assessment (LCA) confirms that H 2 O 2 , NaOH, and energy consumption are the variables with the highest impact from an environmental standpoint. The use of persulfate reduced the relative impact in 1.2 to 12% in 12 of the 18 environmental categories studied using the ReCiPe method. The PFPS process emits 1.23 kg CO 2 (CO 2 -Eqv/m 3 treated water). On the other hand, the PF process emits 1.28 kg CO 2 (CO 2 -Eqv/m 3 treated water). Process intensification, chemometric techniques, and the use of SPRs minimize the impact of some barriers (reagent and energy consumption, technical complexity of reactors, pressure drops, dirt on the reflecting surfaces, fragility of reactor materials), limiting the application of advanced oxidation systems at an industrial level, and decrease treatment cost as well as potential environmental impacts associated with energy and reagents consumption. Treatment costs for PF processes (US$0.78/m 3 ) and PFPS processes (US$0.63/m 3 ) were 20 times lower than those reported for photo-Fenton processes in CPC-type reactors.

Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

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

Tarchalski, M.; Pytel, K.; Wroblewska, M.

2015-07-01

Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to themore » qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from measurements is of the order of 2.5 W/g at half of the possible MARIA power - 15 MW. The approach and the detailed program for experimental verification of calculations will be presented. The following points will be discussed: - Development of a gamma heating model of MARIA reactor with TRIPOLI 4 (coupled neutron-photon mode) and APOLLO2 model taking into account the key parameters like: configuration of the core, experimental loading, control rod location, reactor power, fuel depletion); - Design of specific measurement tools for MARIA experiments including for instance a new single-cell calorimeter called KAROLINA calorimeter; - MARIA experimental program description and a preliminary analysis of results; - Comparison of calculations for JHR and MARIA cores with experimental verification analysis, calculation behavior and n-γ 'environments'. (authors)« less

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

SP-100 program developments

NASA Technical Reports Server (NTRS)

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

1985-01-01

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

48 CFR 242.7400 - General.

Code of Federal Regulations, 2011 CFR

2011-10-01

... Contractor Facilities 242.7400 General. (a) Program managers may conclude that they need technical... manager may assign technical representatives under the procedures in 242.7401. (b) A technical representative is a representative of a DoD program, project, or system office performing non-CAS technical...

48 CFR 242.7400 - General.

Code of Federal Regulations, 2013 CFR

2013-10-01

... Contractor Facilities 242.7400 General. (a) Program managers may conclude that they need technical... manager may assign technical representatives under the procedures in 242.7401. (b) A technical representative is a representative of a DoD program, project, or system office performing non-CAS technical...

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

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

Beatty, Randy L; Harrison, Thomas J

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

A Comparison of Fast-Spectrum and Moderated Space Fission Reactors

NASA Astrophysics Data System (ADS)

Poston, David I.

2005-02-01

The reactor neutron spectrum is one of the fundamental design choices for any fission reactor, but the implications of using a moderated spectrum are vastly different for space reactors as opposed to terrestrial reactors. In addition, the pros and cons of neutron spectra are significantly different among many of the envisioned space power applications. This paper begins with a discussion of the neutronic differences between fast-spectrum and moderated space reactors. This is followed by a discussion of the pros and cons of fast-spectrum and moderated space reactors separated into three areas—technical risk, performance, and safety/safeguards. A mix of quantitative and qualitative arguments is presented, and some conclusions generally can be made regarding neutron spectrum and space power application. In most cases, a fast-spectrum system appears to be the better alternative (mostly because of simplicity and higher potential operating temperatures); however, in some cases, such as a low-power (<100-kWt) surface reactor, a moderated spectrum could provide a better approach. In all cases, the determination of which spectrum is preferred is a strong function of the metrics provided by the "customer"— i.e., if a certain level of performance is required, it could provide a different solution than if a certain level of safeguards is required (which in some cases could produce a null solution). The views expressed in this document are those of the author and do not necessarily reflect agreement by the Government.

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

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

Gerald E. Tripard; Keith G. Fox

2006-08-24

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

An Expert Elicitation of the Proliferation Resistance of Using Small Modular Reactors (SMR) for the Expansion of Civilian Nuclear Systems.

PubMed

Siegel, Jonas; Gilmore, Elisabeth A; Gallagher, Nancy; Fetter, Steve

2018-02-01

To facilitate the use of nuclear energy globally, small modular reactors (SMRs) may represent a viable alternative or complement to large reactor designs. One potential benefit is that SMRs could allow for more proliferation resistant designs, manufacturing arrangements, and fuel-cycle practices at widespread deployment. However, there is limited work evaluating the proliferation resistance of SMRs, and existing proliferation assessment approaches are not well suited for these novel arrangements. Here, we conduct an expert elicitation of the relative proliferation resistance of scenarios for future nuclear energy deployment driven by Generation III+ light-water reactors, fast reactors, or SMRs. Specifically, we construct the scenarios to investigate relevant technical and institutional features that are postulated to enhance the proliferation resistance of SMRs. The experts do not consistently judge the scenario with SMRs to have greater overall proliferation resistance than scenarios that rely on conventional nuclear energy generation options. Further, the experts disagreed on whether incorporating a long-lifetime sealed core into an SMR design would strengthen or weaken proliferation resistance. However, regardless of the type of reactor, the experts judged that proliferation resistance would be enhanced by improving international safeguards and operating several multinational fuel-cycle facilities rather than supporting many more national facilities. © 2017 Society for Risk Analysis.

Workforce Investments: State Strategies to Preserve Higher-Cost Career Education Programs in Community and Technical Colleges

ERIC Educational Resources Information Center

Shulock, Nancy; Lewis, Jodi; Tan, Connie

2013-01-01

In today's highly-skilled economy, rewarding career pathways are available to those who acquire technical skills by enrolling in certificate and associate degree programs in a community or technical college. Such programs are often more costly to offer than liberal arts and sciences programs that prepare students to transfer to four-year…

24 CFR 891.175 - Technical assistance.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Technical assistance. 891.175... Requirements § 891.175 Technical assistance. For purposes of the Section 202 Program and the Section 811 Program, the Secretary shall make available appropriate technical assistance to assure that applicants...

10 CFR 905.40 - Technical assistance.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Technical assistance. 905.40 Section 905.40 Energy DEPARTMENT OF ENERGY ENERGY PLANNING AND MANAGEMENT PROGRAM Energy Services § 905.40 Technical assistance. Western shall establish a program that provides technical assistance to customers to conduct integrated...

STARTUP REACTIVITY ACCOUNTABILITY ATTRIBUTED TO ISOTOPIC TRANSMUTATIONS IN THE IRRADIATED BERYLLIUM REFLECTOR OF THE HIGH FLUX ISTOTOPE REACTOR

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

Chandler, David; Maldonado, G Ivan; Primm, Trent

2010-01-01

The objective of this study is to develop a methodology to predict the reactivity impact as a function of outage time between cycles of 3He, 6Li, and other poisons in the High Flux Isotope Reactor s (HFIR) beryllium reflector. The reactivity worth at startup of the HFIR has been incorrectly predicted in the past after the reactor has been shut-down for long periods of time. The incorrect prediction was postulated to be due to the erroneous calculation of 3He buildup in the beryllium reflector. It is necessary to develop a better estimate of the start-of-cycle symmetric critical control element positionsmore » since if the estimated and actual symmetrical critical control element positions differ by more than $1.55 in reactivity (approximately one-half inch in control element startup position), HFIR is to be shutdown and a technical evaluation is performed to resolve the discrepancy prior to restart. 3He is generated and depleted during operation, but during an outage, the depletion of 3He ceases because it is a stable isotope. 3He is born from the radioactive decay of tritium, and thus the concentration of 3He increases during shutdown. The computer program SCALE, specifically the TRITON and CSAS5 control modules including the KENO V.A, COUPLE, and ORIGEN functional modules were utilized in this study. An equation relating the down time (td) to the change in symmetric control element position was generated and validated against measurements for approximately 40 HFIR operating cycles. The newly-derived correlation was shown to improve accuracy of predictions for long periods of down time.« less

Embedded Sensors and Controls to Improve Component Performance and Reliability -- Loop-scale Testbed Design Report

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

Melin, Alexander M.; Kisner, Roger A.

2016-09-01

Embedded instrumentation and control systems that can operate in extreme environments are challenging to design and operate. Extreme environments limit the options for sensors and actuators and degrade their performance. Because sensors and actuators are necessary for feedback control, these limitations mean that designing embedded instrumentation and control systems for the challenging environments of nuclear reactors requires advanced technical solutions that are not available commercially. This report details the development of testbed that will be used for cross-cutting embedded instrumentation and control research for nuclear power applications. This research is funded by the Department of Energy's Nuclear Energy Enabling Technologymore » program's Advanced Sensors and Instrumentation topic. The design goal of the loop-scale testbed is to build a low temperature pump that utilizes magnetic bearing that will be incorporated into a water loop to test control system performance and self-sensing techniques. Specifically, this testbed will be used to analyze control system performance in response to nonlinear and cross-coupling fluid effects between the shaft axes of motion, rotordynamics and gyroscopic effects, and impeller disturbances. This testbed will also be used to characterize the performance losses when using self-sensing position measurement techniques. Active magnetic bearings are a technology that can reduce failures and maintenance costs in nuclear power plants. They are particularly relevant to liquid salt reactors that operate at high temperatures (700 C). Pumps used in the extreme environment of liquid salt reactors provide many engineering challenges that can be overcome with magnetic bearings and their associated embedded instrumentation and control. This report will give details of the mechanical design and electromagnetic design of the loop-scale embedded instrumentation and control testbed.« less

Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

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

Ross, Kyle W.; Gauntt, Randall O.; Cardoni, Jeffrey N.

2013-11-01

Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.

Highly Selective Nuclide Removal from the R-Reactor Disassembly Basin at the SRS

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

Pickett, J. B.; Austin, W. E.; Dukes, H. H.

This paper describes the results of a deployment of highly selective ion-exchange resin technologies for the in-situ removal of Cs-137 and Sr-90 from the Savannah River Site (SRS) R-Reactor Disassembly Basin. The deployment was supported by the DOE Office of Science and Technology's (OST, EM-50) National Engineering Technology Laboratory (NETL), as a part of an Accelerated Site Technology Deployment (ASTD) project. The Facilities Decontamination and Decommissioning (FDD) Program at the SRS conducted this deployment as a part of an overall program to deactivate three of the site's five reactor disassembly basins.

Highly Selective Nuclide Removal from the R-Reactor Disassembly Basin at SRS

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

Pickett, J.B.

This paper describes the results of a deployment of highly selective ion-exchange resin technologies for the in-situ removal of Cs-137 and Sr-90 from the Savannah River Site (SRS) R-Reactor Disassembly Basin. The deployment was supported by the DOE Office of Science and Technology's (OST, EM-50) National Engineering Technology Laboratory (NETL), as a part of an Accelerated Site Technology Deployment (ASTD) project. The Facilities Decontamination and Decommissioning (FDD) Program at the SRS conducted this deployment as a part of an overall program to deactivate three of the site's five reactor disassembly basins

An approach to model reactor core nodalization for deterministic safety analysis

NASA Astrophysics Data System (ADS)

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

2016-01-01

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

An approach to model reactor core nodalization for deterministic safety analysis

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

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

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

Office for Analysis and Evaluation of Operational Data 1994-FY 95 annual report. Volume 9, Number 2

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

NONE

This annual report of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) describes activities conducted during CY 1994 and FY 1995. The report is published in three parts. NUREG-1272, Vol. 9, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year and summarizes information from such sources as licensee event reports, diagnostic evaluations, and reportsmore » to the NRC`s Operations Center. NUREG-1272, Vol. 9, No. 2, covers nuclear materials and presents a review of the events and concerns associated with the use of licensed material in nonreactor applications, such as personnel overexposures and medical misadministrations. Both reports also contain a discussion of the Incident Investigation Team program and summarize both the Incident Investigation Team and Augmented Inspection Team reports. Each volume contains a list of the AEOD reports issued from 1980 through FY 1995. NUREG-1272, Vol. 9, No. 3, covers technical training and presents the activities of the Technical Training Center in support of the NRC`s mission.« less

Flow Induced Vibration Program at Argonne National Laboratory

NASA Astrophysics Data System (ADS)

1984-01-01

The Argonne National Laboratory's Flow Induced Vibration Program, currently residing in the Laboratory's Components Technology Division is discussed. Throughout its existence, the overall objective of the program was to develop and apply new and/or improved methods of analysis and testing for the design evaluation of nuclear reactor plant components and heat exchange equipment from the standpoint of flow induced vibration. Historically, the majority of the program activities were funded by the US Atomic Energy Commission, the Energy Research and Development Administration, and the Department of Energy. Current DOE funding is from the Breeder Mechanical Component Development Division, Office of Breeder Technology Projects; Energy Conversion and Utilization Technology Program, Office of Energy Systems Research; and Division of Engineering, Mathematical and Geosciences, office of Basic Energy Sciences. Testing of Clinch River Breeder Reactor upper plenum components was funded by the Clinch River Breeder Reactor Plant Project Office. Work was also performed under contract with Foster Wheeler, General Electric, Duke Power Company, US Nuclear Regulatory Commission, and Westinghouse.

Heavy-section steel irradiation program. Semiannual progress report, October 1996--March 1997

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

Rosseel, T.M.

1998-02-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV`s fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established. Its primary goal is to provide a thorough, quantitative assessment of the effects of neutron irradiationmore » on the material behavior and, in particular, the fracture toughness properties of typical pressure-vessel steels as they relate to light-water RPV integrity. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into eight tasks: (1) program management, (2) irradiation effects in engineering materials, (3) annealing, (4) microstructural analysis of radiation effects, (5) in-service irradiated and aged material evaluations, (6) fracture toughness curve shift method, (7) special technical assistance, and (8) foreign research interactions. The work is performed by the Oak Ridge National Laboratory.« less

75 FR 34251 - Office of Special Education and Rehabilitative Services; Overview Information; Centers for...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-16

... Information; Centers for Independent Living Program--Training and Technical Assistance; Notice Inviting... Training and Technical Assistance Project. Program Authority: 29 U.S.C. 796f(b); American Recovery and... Independent Living Program--Training and Technical Assistance--CFDA Number 84.400B must be submitted...

77 FR 16250 - Notice of Proposed Information Collection for Public Comment; Office of Native American Programs...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-20

... Information Collection for Public Comment; Office of Native American Programs (ONAP) Training and Technical... subject proposal. The data required by Office of Native American Programs Training and Technical... progress. The data identifies needs, outputs and outcomes of the training and technical assistance. DATES...

77 FR 9216 - Native American Career and Technical Education Program; Proposed Waivers and Extension of the...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-16

... DEPARTMENT OF EDUCATION Native American Career and Technical Education Program; Proposed Waivers... Education, Department of Education. ACTION: Notice. SUMMARY: For 60-month projects funded in fiscal year (FY) 2007 under the Native American Career and Technical Education Program (NACTEP), the Secretary proposes...

Nuclear Technology Series. Course 6: Instrumentation and Control of Reactors and Plant Systems.

ERIC Educational Resources Information Center

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

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

Proceedings of GLOBAL 2013: International Nuclear Fuel Cycle Conference - Nuclear Energy at a Crossroads

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

NONE

2013-07-01

The Global conference is a forum for the discussion of the scientific, technical, social and regulatory aspects of the nuclear fuel cycle. Relevant topics include global utilization of nuclear energy, current fuel cycle technologies, advanced reactors, advanced fuel cycles, nuclear nonproliferation and public acceptance.

Silicon Chemical Vapor Deposition Process Using a Half-Inch Silicon Wafer for Minimal Manufacturing System

NASA Astrophysics Data System (ADS)

Li, Ning; Habuka, Hitoshi; Ikeda, Shin-ichi; Hara, Shiro

A chemical vapor deposition reactor for producing thin silicon films was designed and developed for achieving a new electronic device production system, the Minimal Manufacturing, using a half-inch wafer. This system requires a rapid process by a small footprint reactor. This was designed and verified by employing the technical issues, such as (i) vertical gas flow, (ii) thermal operation using a highly concentrated infrared flux, and (iii) reactor cleaning by chlorine trifluoride gas. The combination of (i) and (ii) could achieve a low heating power and a fast cooling designed by the heat balance of the small wafer placed at a position outside of the reflector. The cleaning process could be rapid by (iii). The heating step could be skipped because chlorine trifluoride gas was reactive at any temperature higher than room temperature.

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

Goldsmith, M.W.; Forbes, I.A.; Turnage, J.C.

The potential of new and future energy technologies is discussed, with information provided on availability, technical and economic feasibility, and limitations due to the form of the energy. Energy sources not presently in use (i.e., shale oil, garbage, geothermal, wind, tidal, breeder reactors, ocean thermal gradients, solar energy, and fusion) are expected to supply only 10 to 15% of the Nation's energy requirements in the year 2000. The following chapters are included: Energy Use and Supply; Extending Chemical Fuel Resources, which covers oil shale and tar sands, coal gasification and liquefaction, garbage, and biomass energy; Harnessing the Forces of Nature,more » which describes geothermal, tidal, hydro, wind, and solar energy; New Nuclear Technology (e.g., converter reactors, breeder reactors, fusion by magnetic confinement, and laser fusion); and Improving Energy Production Efficiency, with discussions on energy storage, MHD (magnetohydrodynamics), and combined cycles. (64 references) (BYB)« less

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

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

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

2015-07-01

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

10 CFR 72.218 - Termination of licenses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.218 Termination of licenses. (a) The notification regarding the program for the management of spent fuel at the reactor required by § 50.54(bb) of...

10 CFR 72.218 - Termination of licenses.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.218 Termination of licenses. (a) The notification regarding the program for the management of spent fuel at the reactor required by § 50.54(bb) of...

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.

Using materials research results in new regulations -- The Swedish approach

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

Gott, K.

1995-12-31

Swedish regulations are normally divided into two sections: the first part is the compulsory text and the second part explains very briefly the ideas behind the regulations and section consists of an interpretive text. This second part explains very briefly the ideas behind the regulations and gives advice as to how to apply the regulations, acceptable testing and analysis methods, and references to other standards and relevant documents. In the new regulations, which were approved by the Board of SKI in September 1994 and are effective from 1st January 1995, a number of innovations have been included concerning chemistry andmore » environmental degradation of the primary pressure boundary in Light Water Reactors. With regard to chemistry SKI will no longer approve the various parameters in the technical specifications (such as conductivity and impurity concentrations) but will require that the utilities have a chemistry control program in place which ensures the integrity of the primary pressure boundary and does not expose it to environments (such as impurities and decontamination chemicals) for which it was not designed. SKI can at any time control that such a program exists and assess its compatibility with these goals, either during routine inspections or as part of special theme inspections. Crack growth rates have been specified for different materials stainless steels, and the nickel base alloy types 600 and 182. Different environments have also been specified: water chemistry within and outside plant specifications as well as normal and hydrogen water chemistry conditions. Stress corrosion cracking in pressurized water reactor systems is also treated separately in the regulations, but not discussed specifically here.« less

Zirconium Hydride Space Power Reactor design.

NASA Technical Reports Server (NTRS)

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

1972-01-01

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

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

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

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

2014-02-01

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

HYBRID SULFUR PROCESS REFERENCE DESIGN AND COST ANALYSIS

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

Gorensek, M.; Summers, W.; Boltrunis, C.

2009-05-12

This report documents a detailed study to determine the expected efficiency and product costs for producing hydrogen via water-splitting using energy from an advanced nuclear reactor. It was determined that the overall efficiency from nuclear heat to hydrogen is high, and the cost of hydrogen is competitive under a high energy cost scenario. It would require over 40% more nuclear energy to generate an equivalent amount of hydrogen using conventional water-cooled nuclear reactors combined with water electrolysis compared to the proposed plant design described herein. There is a great deal of interest worldwide in reducing dependence on fossil fuels, whilemore » also minimizing the impact of the energy sector on global climate change. One potential opportunity to contribute to this effort is to replace the use of fossil fuels for hydrogen production by the use of water-splitting powered by nuclear energy. Hydrogen production is required for fertilizer (e.g. ammonia) production, oil refining, synfuels production, and other important industrial applications. It is typically produced by reacting natural gas, naphtha or coal with steam, which consumes significant amounts of energy and produces carbon dioxide as a byproduct. In the future, hydrogen could also be used as a transportation fuel, replacing petroleum. New processes are being developed that would permit hydrogen to be produced from water using only heat or a combination of heat and electricity produced by advanced, high temperature nuclear reactors. The U.S. Department of Energy (DOE) is developing these processes under a program known as the Nuclear Hydrogen Initiative (NHI). The Republic of South Africa (RSA) also is interested in developing advanced high temperature nuclear reactors and related chemical processes that could produce hydrogen fuel via water-splitting. This report focuses on the analysis of a nuclear hydrogen production system that combines the Pebble Bed Modular Reactor (PBMR), under development by PBMR (Pty.) Ltd. in the RSA, with the Hybrid Sulfur (HyS) Process, under development by the Savannah River National Laboratory (SRNL) in the US as part of the NHI. This work was performed by SRNL, Westinghouse Electric Company, Shaw, PBMR (Pty) Ltd., and Technology Insights under a Technical Consulting Agreement (TCA). Westinghouse Electric, serving as the lead for the PBMR process heat application team, established a cost-shared TCA with SRNL to prepare an updated HyS thermochemical water-splitting process flowsheet, a nuclear hydrogen plant preconceptual design and a cost estimate, including the cost of hydrogen production. SRNL was funded by DOE under the NHI program, and the Westinghouse team was self-funded. The results of this work are presented in this Final Report. Appendices have been attached to provide a detailed source of information in order to document the work under the TCA contract.« less

FY16 Status Report for the Uranium-Molybdenum Fuel Concept

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

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

2016-09-22

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

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

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

Singleton, A.H.

1995-06-28

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

Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

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

Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.

2011-11-14

For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may bemore » constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.« less

Evaluation of Nuclear Facility Decommissioning Projects program: a reference research reactor. Project summary report

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

Baumann, B.L.; Miller, R.L.

1983-10-01

This document presents, in summary form, generic conceptual information relevant to the decommissioning of a reference research reactor (RRR). All of the data presented were extracted from NUREG/CR-1756 and arranged in a form that will provide a basis for future comparison studies for the Evaluation of Nuclear Facility Decommissioning Projects (ENFDP) program.

Status of Wrought FeCrAl-UO 2 Capsules Irradiated in the Advanced Test Reactor

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

Field, Kevin G.; Harp, J.; Core, G.

2017-07-01

Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction,more » and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.« less

NASA Technical Standards Program and Implications for Lessons Learned and Technical Standard Integration

NASA Technical Reports Server (NTRS)

Gill, Paul S.; Garcia, Danny; Vaughan, William W.; Parker, Nelson C. (Technical Monitor)

2002-01-01

The National Aeronautics and Space Agency consists of fourteen Facilities throughout the United States. They are organized to support the Agency's principal Enterprises: (1) Space Science, (2) Earth Science, (3) Aerospace Technology, (4) Human Exploration and Development of Space, and (5) Biological and Physical Research. Technical Standards are important to the activities of each Enterprise and have been an integral part in the development and operation of NASA Programs and Projects since the Agency was established in 1959. However, for years each Center was responsible for its own standards development and selection of non-NASA technical standards that met the needs of Programs and Projects for which they were responsible. There were few Agencywide applicable Technical Standards, mainly those in area of safety. Department of Defense Standards and Specifications were the foundation and main source for Technical Standards used by the Agency. This process existed until about 1997 when NASA embarked on a Program to convert NASA's Center-developed Technical Standards into Agencywide endorsed NASA Preferred Technical Standards. In addition, action was taken regarding the formal adoption of non-NASA Technical Standards (DOD, SAE, ASTM, ASME, IEEE, etc.) as NASA Preferred Technical Standards.

40 CFR 52.2586 - Small business stationary source technical and environmental compliance assistance program.

Code of Federal Regulations, 2013 CFR

2013-07-01

... PLANS (CONTINUED) Wisconsin § 52.2586 Small business stationary source technical and environmental compliance assistance program. The Wisconsin small business stationary source technical and environmental...

40 CFR 52.2586 - Small business stationary source technical and environmental compliance assistance program.

Code of Federal Regulations, 2010 CFR

2010-07-01

... PLANS (CONTINUED) Wisconsin § 52.2586 Small business stationary source technical and environmental compliance assistance program. The Wisconsin small business stationary source technical and environmental...