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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-02

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

77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-12

... for Passive Advanced Light Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES: Submit comments by November 13, 2012. Comments...

10 CFR 73.4 - Communications.

Code of Federal Regulations, 2012 CFR

2012-01-01

...: Document Control Desk, Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear Material Safety and Safeguards, or Director, Division of Security Policy... 10 Energy 2 2012-01-01 2012-01-01 false Communications. 73.4 Section 73.4 Energy NUCLEAR...

78 FR 20959 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... Committee on Reactor Safeguards. [FR Doc. 2013-08131 Filed 4-5-13; 8:45 am] BILLING CODE 7590-01-P ...

Hanford Works monthly report, December 1950

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

Prout, G.R.

1951-01-22

This is a progress report of the production reactors on the Hanford Reservation for the month of December 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, April 1952

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

Prout, G.R.

1952-05-20

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, August 1950

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

Prout, G.R.

1950-09-18

This is a progress report of the production reactors on the Hanford Reservation for the month of August 1950. This report takes each division (e.g. manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, May 1951

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

Prout, G.R.

1951-06-21

This is a progress report of the production reactors on the Hanford Reservation for the month of May 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, December 1951

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

Prout, G.R.

1952-01-22

This is a progress report of the production reactors on the Hanford Reservation for the month of December 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, March 1951

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

Prout, G.R.

1951-04-20

This is a progress report of the production reactors on the Hanford Reservation for the month of March 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, July 1951

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

Prout, G.R.

This is a progress report of the production reactors on the Hanford Reservation for the month of July 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, March 1952

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

Prout, G.R.

1952-04-18

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, July 1952

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

Prout, G.R.

1952-08-15

This is a progress report of the production reactors on the Hanford Reservation for the month of July 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford works monthly report, September 1951

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

Prout, G.R.

This is a progress report of the production reactors on the Hanford Reservation for the month of September 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, January 1952

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

Prout, G.R.

This is a progress report of the production reactors on the Hanford Reservation for the month of January 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, August 1951

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

Not Available

1951-09-24

This is a progress report of the production reactors on the Hanford Reservation for the month of August 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, July 1950

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

Prout, G.R.

1950-08-18

This is a progress report of the production reactors on the Hanford Reservation for the month of July 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, November 1951

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

Prout, G.R.

1951-12-21

This is a progress report of the production reactors on the Hanford Reservation for the month of November 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, October 1950

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

Prout, G.R.

1950-11-20

This is a progress report of the production reactors on the Hanford Reservation for the month of October 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, September 1950

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

Prout, G.R.

1950-10-20

This is a progress report of the production reactors on the Hanford Reservation for the month of September 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, November 1950

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

Prout, G.R.

1950-12-20

This is a progress report of the production reactors on the Hanford Reservation for the month of November 1950. This report takes each division (e.g. manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

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.

77 FR 76089 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-26

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on January 16, 2013, Room T-2B3, 11545 Rockville Pike...

76 FR 5218 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-28

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on February 8, 2011, 11545 Rockville Pike, Rockville, MD...

French Atomic Energy Commission Decommissioning Programme and Feedback Experience - 12230

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

Guiberteau, Ph.; Nokhamzon, J.G.

Since the French Atomic and Alternatives Energy Commission (CEA) was founded in 1945 to carry out research programmes on use of nuclear, and its application France has set up and run various types of installations: research or prototypes reactors, process study or examination laboratories, pilot installations, accelerators, nuclear power plants and processing facilities. Some of these are currently being dismantled or must be dismantled soon so that the DEN, the Nuclear Energy Division, can construct new equipment and thus have available a range of R and D facilities in line with the issues of the nuclear industry of the future.more » Since the 1960's and 1970's in all its centres, the CEA has acquired experience and know-how through dismantling various nuclear facilities. The dismantling techniques are nowadays operational, even if sometimes certain specific developments are necessary to reduce the cost of operations. Thanks to availability of techniques and guarantees of dismantling programme financing now from two dedicated funds, close to euro 15,000 M for the next thirty years, for current or projected dismantling operations, the CEA's Nuclear Energy Division has been able to develop, when necessary, its immediate dismantling strategy. Currently, nearly thirty facilities are being dismantled by the CEA's Nuclear Energy Division operational units with industrial partners. Thus the next decade will see completion of the dismantling and radioactive clean-up of the Grenoble site and of the facilities on the Fontenay-aux-Roses site. By 2016, the dismantling of the UP1 plant at Marcoule, the largest dismantling work in France, will be well advanced, with all the process equipment dismantled. After an overview of the French regulatory framework, the paper will describe the DD and R (Decontamination Decommissioning and Remediation) strategy, programme and feedback experience inside the CEA's Nuclear Energy Division. A special feature of dismantling operations at the CEA comes from the diversity of facilities to be dismantled, which are predominantly research facilities and therefore have no series advantage. There is tremendous operating feedback, however. For more than twenty years in all its centres, the CEA has acquired experience and know-how through dismantling research reactors or critical models and laboratories or plants. The dismantling techniques are nowadays operational, even if sometimes certain specific developments are necessary to reduce the cost of operations. Thanks to availability of techniques and guarantees of dismantling programme financing from two dedicated funds, close to euro 15,000 Millions for the next thirty years, for current or projected dismantling operations, the Nuclear Energy Division has been able to develop, when necessary, its immediate dismantling strategy. Currently, nearly thirty facilities are being dismantled by the CEA's Nuclear Energy Division operational units with industrial partners. Thus the next decade will see completion of the dismantling and radioactive clean-up of the Grenoble site and of the facilities on the Fontenay-aux-Roses site. By 2020, the dismantling of the UP1 plant at Marcoule, one of the largest dismantling works in the world, will be well advanced, with all the process equipment dismantled. (authors)« less

11. Photocopy of drawing, February 1958. WATERTOWN ARSENAL REACTOR, SHIELD ...

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

11. Photocopy of drawing, February 1958. WATERTOWN ARSENAL REACTOR, SHIELD STRUCTURE, SECTIONS, AND PLANS. Bendix Aviation Corporation; and Giffels & Vallet, Inc., L. Rosetti, Associated Architects and Engineers, Detroit, Michigan. Drawing Number 53-198. (Original: AMTL Engineering Division, Watertown). - Watertown Arsenal, Building No. 100, Wooley Avenue, Watertown, Middlesex County, MA

Hanford Atomic Products Operation monthly report, April 1954

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

McCune, F.K.

1954-05-21

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1954. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Agile Port and High Speed Ship Technologies, Vol 1: FY05 Projects 3-6 and 8-10

DTIC Science & Technology

2008-07-02

Computational Fluid Dynamics DTMB - David Taylor Model Basin JVR - Jet Velocity Ratio NSWCCD - Naval Surface Warfare Center, Carderock Division SDD - Systems...immature current state of the technology employed for the reactor system (multiple closed Brayton Cycle, Helium Cooled Gas reactors); (iii) several

Hanford Atomic Products Operation monthly report, March 1953

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

McCune, F.K.

1953-04-22

This is a progress report of the production reactors on the Hanford Reservation for the month of March 1953. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Atomic Products Operation monthly report, April 1953

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

McCune, F.K.

1953-05-20

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Atomic Products Operation monthly report, January 1954

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

McCune, F.K.

1954-02-25

This is a progress report of the production reactors on the Hanford Reservation for the month of January 1954. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes the accomplishments and employee relations for that month.

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.

Advanced Test Reactor Tour

ScienceCinema

Miley, Don

2017-12-21

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

77 FR 59678 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-28

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on October 2, 2012, Room T-2B1, 11545 Rockville Pike, Rockville...

76 FR 34276 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-13

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on June 21, 2011, Room T-2B1, 11545 Rockville Pike, Rockville...

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2010 CFR

2010-01-01

... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2011 CFR

2011-01-01

... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

Solid State Division progress report for period ending March 31, 1992

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

Green, P.H.; Hinton, L.W.

1992-09-01

During this period, the division conducted a broad, interdisciplinary materials research program with emphasis on theoretical solid state physics, superconductivity, neutron scattering, synthesis and characterization of materials, ion beam and laser processing, and the structure of solids and surfaces. The High Flux Isotope Reactor was returned to full operation.

78 FR 66968 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-07

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on November 22, 2013, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The meeting will be...

78 FR 37595 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting on July 9, 2013, Room T-2B3, 11545 Rockville Pike, Rockville, Maryland. The meeting will be open to...

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

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

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

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

Wichman, K.; Tsao, J.; Mayfield, M.

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRCmore » and the International Piping Integrity Research Group is also briefly summarized.« less

76 FR 62095 - Exelon Generation Company, LLC; Notice of Withdrawal of Application for Amendment to Facility...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-06

... Licensing Branch III-2, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-373 and 50-374; NRC-2011-0234] Exelon Generation.... Nuclear Regulatory Commission (NRC, the Commission) has granted the request of Exelon Generation Company...

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

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

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

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

Cladding and duct materials for advanced nuclear recycle reactors

NASA Astrophysics Data System (ADS)

Allen, T. R.; Busby, J. T.; Klueh, R. L.; Maloy, S. A.; Toloczko, M. B.

2008-01-01

The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.

Nuclear Fuels & Materials Spotlight Volume 5

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

Petti, David Andrew

2016-10-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system.more » • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.« less

75 FR 45158 - Holcim (US) Inc. Corporate Division Including On-Site Leased Workers From Manpower, Office Team...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-02

.... Corporate Division Including On-Site Leased Workers From Manpower, Office Team and Advance Temporary... from Manpower and Office Team, Dundee, Michigan. The notice was published in the Federal Register on... Holcim (US) Inc., Corporate Division, including on-site leased workers from Manpower, Office Team Advance...

Advanced Demonstration and Test Reactor Options Study

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

Petti, David Andrew; Hill, R.; Gehin, J.

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercializationmore » of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy”. Advanced reactors are defined in this study as reactors that use coolants other than water. Advanced reactor technologies have the potential to expand the energy applications, enhance the competitiveness, and improve the sustainability of nuclear energy.« less

Advanced Reactor Technology/Energy Conversion Project FY17 Accomplishments.

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

Rochau, Gary E.

The purpose of the ART Energy Conversion (EC) Project is to provide solutions to convert the heat from an advanced reactor to useful products that support commercial application of the reactor designs.

AN EVALUATION OF HEAVY WATER REACTORS FOR POWER

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

Herron, D.P.; Newkirk, W.H.; Puishes, A.

1957-10-01

Reference designs for pressurized and direct-boiling D/sub 2/O reactors were prepared for electrical outputs of 20, 100, and 250 electrical Mw. A number of possible core designs were considered and those utilized which seemed most appropriate to give low-cost power. The technology and costs available today were employed in the preparation of the over-all plant designs. The Consolidated Western Steel Division of U. S. Steel Corporation assisted by preparing a comprehensive report on the design of large pressure vessels and containment vessels. Zr-clad U fuel elements were used as the study basis, but the effect of using UO/sub 2/ andmore » stainless steel cladding was also considered. The principal results found were: (1) Over a wide range of operating conditions snd economic situations, enriched U (up to perhaps 1.4% U/sup 235/) is presently more economic to employ in D/sub 2/O reactors than is natural U. (2) In the longer range, the use of natural U may become more economic as Zr fabrication costs decrease, continuous charge-discharge devices are developed to permit longer exposure levels, and pressure-vessel technology advances so that the large critical masses and core diameters required are not such sn economic penalty on the natural U. The results agree quite well with the data and discussions of the Canadians. (auth)« less

Warthog: A MOOSE-Based Application for the Direct Code Coupling of BISON and PROTEUS

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

McCaskey, Alexander J.; Slattery, Stuart; Billings, Jay Jay

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Department of Energy's Office of Nuclear Energy provides a robust toolkit for the modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines: two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line and the Reactor Product line provide advanced computational technologies that serve each respective field well, however, their current lack of integration presents a major impediment to future improvements of simulation solution fidelity. Theremore » is a desire for the capability to mix and match tools across Product Lines in an effort to utilize the best from both to improve NEAMS modeling and simulation technologies. This report details a new effort to provide this Product Line interoperability through the development of a new application called Warthog. This application couples the BISON Fuel Performance application from the Fuels Product Line and the PROTEUS Core Neutronics application from the Reactors Product Line in an effort to utilize the best from all parts of the NEAMS toolkit and improve overall solution fidelity of nuclear fuel simulations. To achieve this, Warthog leverages as much prior work from the NEAMS program as possible, and in doing so, enables interoperability between the disparate MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. This report describes this work in full. We begin with a detailed look at the individual NEAMS framework technologies used and developed in the various Product Lines, and the current status of their interoperability. We then introduce the Warthog application: its overall architecture and the ways it leverages the best existing tools from across the NEAMS toolkit to enable BISON-PROTEUS integration. Furthermore, we show how Warthog leverages a tool known as DataTransferKit to seamlessly enable the transfer for solution data between disparate frameworks and mesh formats. To end, we demonstrate tests for the direct software coupling of BISON and PROTEUS using Warthog, and discuss current impediments and solutions to the construction of physically realistic input models for this coupled BISON-PROTEUS system.« less

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

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

NONE

1993-09-15

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

NTRE extended life feasibility assessment

NASA Technical Reports Server (NTRS)

1993-01-01

Results of a feasibility analysis of a long life, reusable nuclear thermal rocket engine are presented in text and graph form. Two engine/reactor concepts are addressed: the Particle Bed Reactor (PBR) design and the Commonwealth of Independent States (CIS) concept. Engine design, integration, reliability, and safety are addressed by various members of the NTRE team from Aerojet Propulsion Division, Energopool (Russia), and Babcock & Wilcox.

75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-09

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on ABWR will... would result in major inconvenience. Dated: March 3, 2010. Antonio F. Dias, Chief, Reactor Safety Branch...

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

Payment of Advanced Placement Exam Fees by Virginia Public School Divisions and Its Impact on Advanced Placement Enrollment and Scores

ERIC Educational Resources Information Center

Cirillo, Mary Grupe

2010-01-01

The purpose of this study was to determine the impact of Virginia school divisions' policy of paying the fee for students to take Advanced Placement exams on Advanced Placement course enrollment, the number of Advanced Placement exams taken by students, the average scores earned and the percent of students earning qualifying scores of 3, 4, or 5…

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

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

Not Available

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Advance notification of shipment of irradiated reactor... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... required under this section for shipments of irradiated reactor fuel in quantities less than that subject...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Advance notification of shipment of irradiated reactor... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... required under this section for shipments of irradiated reactor fuel in quantities less than that subject...

CHEMICAL ENGINEERING DIVISION SUMMARY REPORT, OCTOBER, NOVEMBER, DECEMBER 1960

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

None

1961-03-01

Chemical-metallurgical processing studies were made of pyrometallurgical development snd research, and fuel processing facilities for EBR-II. Fuel-cycle applications of fluidization and volatility techniques included laboratory investigations of fluoride volatility processes, engineeringscale development, and conversion of UF/sub 6/ to UO/sub 2/. Reactor safety studies consisted of metal oxidation and ignition kinetics, and metal-water reactions. Reactor chemistry investigations were conducted to determine nuclear constants and suitable reactor decontamination methods. Routine operations are summarized for the high-level gammairradiation facillty and waste processing. (B.O.G.)

ETR, TRA642. BASEMENT SPACE ALLOCATION FOR EXPERIMENTERS CA. 1966, SOUTHEAST ...

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

ETR, TRA-642. BASEMENT SPACE ALLOCATION FOR EXPERIMENTERS CA. 1966, SOUTHEAST QUADRANT OF FLOOR. WESTINGHOUSE ATOMIC POWER DIVISION (WAPD) AND BETTIS ATOMIC POWER LABORATORY (BAPL) CONSUME MOST OF THE QUADRANT. PHILLIPS PETROLEUM COMPANY ETR-E-2256, 12/1966. INL INDEX NO. 532-0642-00-706-021256, REV. F. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

77 FR 36581 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-19

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S. Advanced Pressurized Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Advanced Pressurized Power Reactor (US-APWR) will hold a meeting on July 9-10, 2012, Room T-2B3, 11545...

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

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)

ANALYTICAL CHEMISTRY DIVISION ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31, 1961

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

None

1962-02-01

Research and development progress is reported on analytlcal instrumentation, dlssolver-solution analyses, special research problems, reactor projects analyses, x-ray and spectrochemical analyses, mass spectrometry, optical and electron microscopy, radiochemical analyses, nuclear analyses, inorganic preparations, organic preparations, ionic analyses, infrared spectral studies, anodization of sector coils for the Analog II Cyclotron, quality control, process analyses, and the Thermal Breeder Reactor Projects Analytical Chemistry Laboratory. (M.C.G.)

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

Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

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

Loflin, Leonard; McRimmon, Beth

2014-12-18

This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

Instrumentation and Controls Division Overview: Sensors Development for Harsh Environments at Glenn Research Center

NASA Technical Reports Server (NTRS)

Zeller, Mary V.; Lei, Jih-Fen

2002-01-01

The Instrumentation and Controls Division is responsible for planning, conducting and directing basic and applied research on advanced instrumentation and controls technologies for aerospace propulsion and power applications. The Division's advanced research in harsh environment sensors, high temperature high power electronics, MEMS (microelectromechanical systems), nanotechnology, high data rate optical instrumentation, active and intelligent controls, and health monitoring and management will enable self-feeling, self-thinking, self-reconfiguring and self-healing Aerospace Propulsion Systems. These research areas address Agency challenges to deliver aerospace systems with reduced size and weight, and increased functionality and intelligence for future NASA missions in advanced aeronautics, economical space transportation, and pioneering space exploration. The Division also actively supports educational and technology transfer activities aimed at benefiting all humankind.

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

Communications and Intelligent Systems Division - Division Overview

NASA Technical Reports Server (NTRS)

Miranda, Felix A.

2017-01-01

This presentation provides an overview of the research and engineering work being performed in the competency fields of advanced communications and intelligent systems with emphasis on advanced technologies, architecture definition,and systems development for application in current and future aeronautics and space communications systems.

Communications and Intelligent Systems Division - Division Overview

NASA Technical Reports Server (NTRS)

Miranda, Felix A.

2017-01-01

This presentation provides an overview of the research and engineering work being performed in the competency fields of advanced communications and intelligent systems with emphasis on advanced technologies, architecture definition, and systems development for application in current and future aeronautics and space communications systems.

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

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

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

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

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...

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

Advanced Plasma Pyrolysis Assembly (PPA) Reactor and Process Development

NASA Technical Reports Server (NTRS)

Wheeler, Richard R., Jr.; Hadley, Neal M.; Dahl, Roger W.; Abney, Morgan B.; Greenwood, Zachary; Miller, Lee; Medlen, Amber

2012-01-01

Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA's Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development.

Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors

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

Fleischman, R.M.; Goldsmith, S.; Newman, D.F.

1981-09-01

The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are toomore » costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized.« less

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

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

Conceptual Assessment Tool for Advanced Undergraduate Electrodynamics

ERIC Educational Resources Information Center

Baily, Charles; Ryan, Qing X.; Astolfi, Cecilia; Pollock, Steven J.

2017-01-01

As part of ongoing investigations into student learning in advanced undergraduate courses, we have developed a conceptual assessment tool for upper-division electrodynamics (E&M II): the Colorado UppeR-division ElectrodyNamics Test (CURrENT). This is a free response, postinstruction diagnostic with 6 multipart questions, an optional 3-question…

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

Jurisson, Silvia S.; Lever, Susan Z.; Robertson, J. David

This grant was situated at the University of Missouri to train Ph.D. scientists in radiochemistry and synthetic chemistry in conjunction with Faculty from the Interdisciplinary Plant Group, Division of Biological Sciences, the MU Research Reactor Center, Molecular Biology and the Radiopharmaceutical Sciences Institute. This project was collaborative with Brookhaven National Laboratory (Richard Ferrieri, PI). Projects for the Ph.D. candidates included novel probe development for peptides, nucleosides, small molecules or radiometals, the direct use of radiometals as probes, or nuclear techniques for analysis. The projects for the postdoctoral fellow involved synthetic chemistry for the preparation of precursors for novel tracers thatmore » will be radiolabeled with 18F or other appropriate radionuclides. The skill sets of our team members allowed us to prepare probes with positron or single photon emitters, as well as ones that are dual-labeled (fluorescent and radiolabeled). We focused our technical advances to those that will be broadly applicable to any research field.« less

Study of a Multi-Phase Hybrid Heat Exchanger-Reactor (HEX Reactor): Part 2 - Numerical Prediction of Thermal Performance (Postprint)

DTIC Science & Technology

2014-01-01

PERSON (Monitor) a. REPORT Unclassified b . ABSTRACT Unclassified c. THIS PAGE Unclassified Travis E. Michalak 19b. TELEPHONE NUMBER (Include...distribution unlimited.Nicholas Niedbalski a, b ,⇑, Douglas Johnson c, Soumya S. Patnaik a, Debjyoti Banerjee b aAir Force Research Laboratory, Aerospace...Systems Directorate, Power and Controls Division, Mechanical and Thermal Systems Branch, 1950 5th St., Wright-Patterson AFB, OH 45433, United States b Texas

Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

PubMed

Hill, R N; Nutt, W M; Laidler, J J

2011-01-01

The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. Copyright © 2010 Health Physics Society

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

Tan, Lizhen; Yang, Ying; Chen, Tianyi

Advanced nuclear reactors as well as the life extension of light water reactors require advanced alloys capable of satisfactory operation up to neutron damage levels approaching 200 displacements per atom (dpa). Extensive studies, including fundamental theories, have demonstrated the superior resistance to radiation-induced swelling in ferritic steels, primarily inherited from their body-centered cubic (bcc) structure. This study aims at developing nanoprecipitates strengthened advanced ferritic alloys for advanced nuclear reactor applications. To be more specific, this study aims at enhancing the amorphization ability of some precipitates, such as Laves phase and other types of intermetallic phases, through smart alloying strategy, andmore » thereby promote the crystalline®amorphous transformation of these precipitates under irradiation.« less

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

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

NASA Astrophysics Data System (ADS)

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

1981-09-01

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

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

NASA Technical Reports Server (NTRS)

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

1981-01-01

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

Modular fabrication and characterization of complex silicon carbide composite structures Advanced Reactor Technologies (ART) Research Final Report (Feb 2015 – May 2017)

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

Khalifa, Hesham

Advanced ceramic materials exhibit properties that enable safety and fuel cycle efficiency improvements in advanced nuclear reactors. In order to fully exploit these desirable properties, new processing techniques are required to produce the complex geometries inherent to nuclear fuel assemblies and support structures. Through this project, the state of complex SiC-SiC composite fabrication for nuclear components has advanced significantly. New methods to produce complex SiC-SiC composite structures have been demonstrated in the form factors needed for in-core structural components in advanced high temperature nuclear reactors. Advanced characterization techniques have been employed to demonstrate that these complex SiC-SiC composite structures providemore » the strength, toughness and hermeticity required for service in harsh reactor conditions. The complex structures produced in this project represent a significant step forward in leveraging the excellent high temperature strength, resistance to neutron induced damage, and low neutron cross section of silicon carbide in nuclear applications.« less

Development of a Polysilicon Process Based on Chemical Vapor Deposition of Dichlorosilane in an Advanced Siemen's Reactor

NASA Technical Reports Server (NTRS)

Arevidson, A. N.; Sawyer, D. H.; Muller, D. M.

1983-01-01

Dichlorosilane (DCS) was used as the feedstock for an advanced decomposition reactor for silicon production. The advanced reactor had a cool bell jar wall temperature, 300 C, when compared to Siemen's reactors previously used for DCS decomposition. Previous reactors had bell jar wall temperatures of approximately 750 C. The cooler wall temperature allows higher DCS flow rates and concentrations. A silicon deposition rate of 2.28 gm/hr-cm was achieved with power consumption of 59 kWh/kg. Interpretation of data suggests that a 2.8 gm/hr-cm deposition rate is possible. Screening of lower cost materials of construction was done as a separate program segment. Stainless Steel (304 and 316), Hastalloy B, Monel 400 and 1010-Carbon Steel were placed individually in an experimental scale reactor. Silicon was deposited from trichlorosilane feedstock. The resultant silicon was analyzed for electrically active and metallic impurities as well as carbon. No material contributed significant amounts of electrically active or metallic impurities, but all contributed carbon.

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

Plant maintenance and advanced reactors, 2005

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

Agnihotri, Newal

2005-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: First U.S. EPRs in 2015, by Ray Ganthner, Framatome ANP; Pursuing several opportunities, by William E. (Ed) Cummins, Westinghouse Electric Company; Vigorous plans to develop advanced reactors, by Yuliang Sun, Tsinghua University, China; Multiple designs, small and large, by Kumiaki Moriya, Hitachi Ltd., Japan; Sealed and embedded for safety and security, by Handa Norihiko, Toshiba Corporation, Japan; Scheduled online in 2010, by Johan Slabber, PMBR (Pty) Ltd., South Africa; Multi-application reactors, by Nikolay G. Kodochigov, OKBM, Russia; Six projects under budgetmore » and on schedule, by David F. Togerson, AECL, Canada; Creating a positive image, by Scott Peterson, Nuclear Energy Institute (NEI); Advanced plans for nuclear power's renaissance, by John Cleveland, International Atomic Energy Agency, Austria; and, Plant profile: last five outages in less than 20 days, by Beth Rapczynski, Exelon Nuclear.« less

77 FR 21592 - Guidelines for Preparing and Reviewing Licensing Applications for the Production of Radioisotopes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-10

... Regulatory Commission. Jessie F. Quichocho, Chief, Research and Test Reactors Licensing Branch, Division of... information in comment submissions that you do not want to be publicly disclosed. The NRC posts all comment...

IECEC '83; Proceedings of the Eighteenth Intersociety Energy Conversion Engineering Conference, Orlando, FL, August 21-26, 1983. Volume 1 - Thermal energy systems

NASA Astrophysics Data System (ADS)

Among the topics discussed are the nuclear fuel cycle, advanced nuclear reactor designs, developments in central status power reactors, space nuclear reactors, magnetohydrodynamic devices, thermionic devices, thermoelectric devices, geothermal systems, solar thermal energy conversion systems, ocean thermal energy conversion (OTEC) developments, and advanced energy conversion concepts. Among the specific questions covered under these topic headings are a design concept for an advanced light water breeder reactor, energy conversion in MW-sized space power systems, directionally solidified cermet electrodes for thermionic energy converters, boron-based high temperature thermoelectric materials, geothermal energy commercialization, solar Stirling cycle power conversion, and OTEC production of methanol. For individual items see A84-30027 to A84-30055

Generating unstructured nuclear reactor core meshes in parallel

DOE PAGES

Jain, Rajeev; Tautges, Timothy J.

2014-10-24

Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

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

NONE

The Chemical Technology (CMT) Division is a diverse technical organization with principal emphases in environmental management and development of advanced energy sources. The Division conducts research and development in three general areas: (1) development of advanced power sources for stationary and transportation applications and for consumer electronics, (2) management of high-level and low-level nuclear wastes and hazardous wastes, and (3) electrometallurgical treatment of spent nuclear fuel. The Division also performs basic research in catalytic chemistry involving molecular energy resources, mechanisms of ion transport in lithium battery electrolytes, and the chemistry of technology-relevant materials and electrified interfaces. In addition, the Divisionmore » operates the Analytical Chemistry Laboratory, which conducts research in analytical chemistry and provides analytical services for programs at Argonne National Laboratory (ANL) and other organizations. Technical highlights of the Division`s activities during 1997 are presented.« less

Eddy Current Flow Measurements in the FFTF

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

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

2017-02-02

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

Nuclear Reactor Physics

NASA Astrophysics Data System (ADS)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

Warthog: Progress on Coupling BISON and PROTEUS

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

Hart, Shane W.D.

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Office of Nuclear Energy at the Department of Energy (DOE) provides a robust toolkit for modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines, with two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line (FPL) and the Reactor Product Line (RPL) provide advanced computational technologies that serve each respective field effectively. There is currently a lack of integration between the product lines, impeding futuremore » improvements of simulation solution fidelity. In order to mix and match tools across the product lines, a new application called Warthog was produced. Warthog is built on the Multi-physics Object-Oriented Simulation Environment (MOOSE) framework developed at Idaho National Laboratory (INL). This report details the continuing efforts to provide the Integration Product Line (IPL) with interoperability using the Warthog code. Currently, this application strives to couple the BISON fuel performance application from the FPL using the PROTEUS Core Neutronics application from the RPL. Warthog leverages as much prior work from the NEAMS program as possible, enabling interoperability between the independently developed MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. Previous work performed on Warthog allowed it to couple a pin cell between the two codes. However, as the temperature changed due to the BISON calculation, the cross sections were not recalculated, leading to errors as the temperature got further away from the initial conditions. XSProc from the SCALE code suite was used to calculate the cross sections as needed. The remainder of this report discusses the changes to Warthog to allow for the implementation of XSProc as an external code. It also discusses the changes made to Warthog to allow it to fit more cleanly into the MultiApp syntax of the MOOSE framework. The capabilities, design, and limitations of Warthog will be described, in addition to some of the test cases that were used to demonstrate the code. Future plans for Warthog will be discussed, including continuation of the modifications to the input and coupling to other SHARP codes such as Nek5000.« less

Regulatory Risk Reduction for Advanced Reactor Technologies – FY2016 Status and Work Plan Summary

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

Moe, Wayne Leland

2016-08-01

Millions of public and private sector dollars have been invested over recent decades to realize greater efficiency, reliability, and the inherent and passive safety offered by advanced nuclear reactor technologies. However, a major challenge in experiencing those benefits resides in the existing U.S. regulatory framework. This framework governs all commercial nuclear plant construction, operations, and safety issues and is highly large light water reactor (LWR) technology centric. The framework must be modernized to effectively deal with non-LWR advanced designs if those designs are to become part of the U.S energy supply. The U.S. Department of Energy’s (DOE) Advanced Reactor Technologiesmore » (ART) Regulatory Risk Reduction (RRR) initiative, managed by the Regulatory Affairs Department at the Idaho National Laboratory (INL), is establishing a capability that can systematically retire extraneous licensing risks associated with regulatory framework incompatibilities. This capability proposes to rely heavily on the perspectives of the affected regulated community (i.e., commercial advanced reactor designers/vendors and prospective owner/operators) yet remain tuned to assuring public safety and acceptability by regulators responsible for license issuance. The extent to which broad industry perspectives are being incorporated into the proposed framework makes this initiative unique and of potential benefit to all future domestic non-LWR applicants« less

Assessment of Sensor Technologies for Advanced Reactors

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

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

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

LWRS ATR Irradiation Testing Readiness Status

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

Kristine Barrett

2012-09-01

The Light Water Reactor Sustainability (LWRS) Program was established by the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. The LWRS Program is divided into four R&D Pathways: (1) Materials Aging and Degradation; (2) Advanced Light Water Reactor Nuclear Fuels; (3) Advanced Instrumentation, Information and Control Systems; and (4) Risk-Informed Safety Margin Characterization. This report describes an irradiation testing readiness analysis in preparation of LWRS experiments for irradiation testing at the Idaho National Laboratory (INL) Advanced Testmore » Reactor (ATR) under Pathway (2). The focus of the Advanced LWR Nuclear Fuels Pathway is to improve the scientific knowledge basis for understanding and predicting fundamental performance of advanced nuclear fuel and cladding in nuclear power plants during both nominal and off-nominal conditions. This information will be applied in the design and development of high-performance, high burn-up fuels with improved safety, cladding integrity, and improved nuclear fuel cycle economics« less

Neutron fluxes in test reactors

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

Youinou, Gilles Jean-Michel

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

Progress report No. 2, design of new pile area {open_quotes}G,{close_quotes} Project C-300, Atomic Energy Commission Directive HW-104

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

Lowe, P.E.

1950-01-16

This document is the progress report for the design and development of Pile Area {open_quotes}G{close_quotes}, to cover the period June 1, 1949 to December 31, 1949. This project represents the major effort of the Reactor Division of the Design and Construction Divisions. It is being pursued with the aim of being able to incorporate the major project objectives in a finished pile design if that design is begun about January, 1951.

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.

A roadmap for nuclear energy technology

NASA Astrophysics Data System (ADS)

Sofu, Tanju

2018-01-01

The prospects for the future use of nuclear energy worldwide can best be understood within the context of global population growth, urbanization, rising energy need and associated pollution concerns. As the world continues to urbanize, sustainable development challenges are expected to be concentrated in cities of the lower-middle-income countries where the pace of urbanization is fastest. As these countries continue their trajectory of economic development, their energy need will also outpace their population growth adding to the increased demand for electricity. OECD IEA's energy system deployment pathway foresees doubling of the current global nuclear capacity by 2050 to reduce the impact of rapid urbanization. The pending "retirement cliff" of the existing U.S. nuclear fleet, representing over 60 percent of the nation's emission-free electricity, also poses a large economic and environmental challenge. To meet the challenge, the U.S. DOE has developed the vision and strategy for development and deployment of advanced reactors. As part of that vision, the U.S. government pursues programs that aim to expand the use of nuclear power by supporting sustainability of the existing nuclear fleet, deploying new water-cooled large and small modular reactors to enable nuclear energy to help meet the energy security and climate change goals, conducting R&D for advanced reactor technologies with alternative coolants, and developing sustainable nuclear fuel cycle strategies. Since the current path relying heavily on water-cooled reactors and "once-through" fuel cycle is not sustainable, next generation nuclear energy systems under consideration aim for significant advances over existing and evolutionary water-cooled reactors. Among the spectrum of advanced reactor options, closed-fuel-cycle systems using reactors with fast-neutron spectrum to meet the sustainability goals offer the most attractive alternatives. However, unless the new public-private partnership models emerge to tackle the licensing and demonstration challenges for these advanced reactor concepts, realization of their enormous potential is not likely, at least in the U.S.

75 FR 8154 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-23

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards In accordance with the... on Reactor Safeguards (ACRS) will hold a meeting on March 4-6, 2010, 11545 Rockville Pike, Rockville....-12 p.m.: New Advanced Reactor Designs (Open)--The Committee will hear presentations by and hold...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-23

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

Physical and chemical controls on the critical zone

USGS Publications Warehouse

Anderson, S.P.; Von Blanckenburg, F.; White, A.F.

2007-01-01

Geochemists have long recognized a correlation between rates of physical denudation and chemical weathering. What underlies this correlation? The Critical Zone can be considered as a feed-through reactor. Downward advance of the weathering front brings unweathered rock into the reactor. Fluids are supplied through precipitation. The reactor is stirred at the top by biological and physical processes. The balance between advance of the weathering front by mechanical and chemical processes and mass loss by denudation fixes the thickness of the Critical Zone reactor. The internal structure of this reactor is controlled by physical processes that create surface area, determine flow paths, and set the residence time of material in the Critical Zone. All of these impact chemical weathering flux.

Background Perchlorate Source Identification Technical Guidance

DTIC Science & Technology

2013-12-01

Sciences Branch (Code 71752) of the Advanced Systems and Applied Sciences Division (Code 71700), Space and Naval Warfare Systems Center (SSC Pacific), San...Head Advanced Systems & Applied Sciences Division iii EXECUTIVE SUMMARY The objective of this document is to outline the approach, tools, and...Helium HMX Octahydro-1,3,5,7-Tetranitro-1,3,5,7- Tetrazocine IR Installation Restoration IRIS Integrated Risk Information System IR-MS Isotope-Ratio

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

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

77 FR 64563 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on November 1-3, 2012, 11545 Rockville...-Term Core Cooling Approach for the Advanced Boiling Water Reactor (ABWR) Design for South Texas Project...

75 FR 55365 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Joint Subcommittee

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-10

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Joint Subcommittee The ACRS Subcommittees on Thermal Hydraulics Phenomena; Advanced Boiling Water Reactor (ABWR); and Materials, Metallurgy, and Reactor Fuels will hold a joint meeting on October 4, 2010...

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.

75 FR 9626 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-03

... maintain variables and systems that can affect the fission process, the integrity of the reactor core, the... on some other defined basis. GDC 24 requires that interconnection of the protection and control..., Regulatory Guide Development Branch, Division of Engineering, Office of Nuclear Regulatory Research. [FR Doc...

Advances in the NASA Earth Science Division Applied Science Program

NASA Astrophysics Data System (ADS)

Friedl, L.; Bonniksen, C. K.; Escobar, V. M.

2016-12-01

The NASA Earth Science Division's Applied Science Program advances the understanding of and ability to used remote sensing data in support of socio-economic needs. The integration of socio-economic considerations in to NASA Earth Science projects has advanced significantly. The large variety of acquisition methods used has required innovative implementation options. The integration of application themes and the implementation of application science activities in flight project is continuing to evolve. The creation of the recently released Earth Science Division, Directive on Project Applications Program and the addition of an application science requirement in the recent EVM-2 solicitation document NASA's current intent. Continuing improvement in the Earth Science Applications Science Program are expected in the areas of thematic integration, Project Applications Program tailoring for Class D missions and transfer of knowledge between scientists and projects.

Investigation of Alien Wavelength Quality in Live Multi-Domain, Multi-Vendor Link Using Advanced Simulation Tool

NASA Astrophysics Data System (ADS)

Nordal Petersen, Martin; Nuijts, Roeland; Lange Bjørn, Lars

2014-05-01

This article presents an advanced optical model for simulation of alien wavelengths in multi-domain and multi-vendor dense wavelength-division multiplexing networks. The model aids optical network planners with a better understanding of the non-linear effects present in dense wavelength-division multiplexing systems and better utilization of alien wavelengths in future applications. The limiting physical effects for alien wavelengths are investigated in relation to power levels, channel spacing, and other factors. The simulation results are verified through experimental setup in live multi-domain dense wavelength-division multiplexing systems between two national research networks: SURFnet in Holland and NORDUnet in Denmark.

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

Kristine Barrett; Shannon Bragg-Sitton

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system thatmore » would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.« 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

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

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

None, None

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

Johnson Noise Thermometry for Advanced Small Modular Reactors

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

Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.

Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor’s physical condition. In and near the core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurementmore » due to their fundamental natures. Small Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of the current ORNL-led project, conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced SMR Research and Development (R&D) program, is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.« less

2015 Groundwater Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Ponds

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

Lewis, Michael George

This report summarizes radiological monitoring results from groundwater wells associated with the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Ponds Reuse Permit (I-161-02). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

75 FR 71744 - Toshiba Corporation Power Systems Company Notice of Receipt and Availability of an Application...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-24

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0361] Toshiba Corporation Power Systems Company Notice of Receipt and Availability of an Application for Renewal of the U.S. Advanced Boiling Water Reactor Design... application for a design certification (DC) renewal for the U.S. Advanced Boiling Water Reactor (ABWR). An...

Membrane contactor/separator for an advanced ozone membrane reactor for treatment of recalcitrant organic pollutants in water

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

Chan, Wai Kit, E-mail: kekyeung@ust.hk; Joueet, Justine; Heng, Samuel

An advanced ozone membrane reactor that synergistically combines membrane distributor for ozone gas, membrane contactor for pollutant adsorption and reaction, and membrane separator for clean water production is described. The membrane reactor represents an order of magnitude improvement over traditional semibatch reactor design and is capable of complete conversion of recalcitrant endocrine disrupting compounds (EDCs) in water at less than three minutes residence time. Coating the membrane contactor with alumina and hydrotalcite (Mg/Al=3) adsorbs and traps the organics in the reaction zone resulting in 30% increase of total organic carbon (TOC) removal. Large surface area coating that diffuses surface chargesmore » from adsorbed polar organic molecules is preferred as it reduces membrane polarization that is detrimental to separation. - Graphical abstract: Advanced ozone membrane reactor synergistically combines membrane distributor for ozone, membrane contactor for sorption and reaction and membrane separator for clean water production to achieve an order of magnitude enhancement in treatment performance compared to traditional ozone reactor. Highlights: Black-Right-Pointing-Pointer Novel reactor using membranes for ozone distributor, reaction contactor and water separator. Black-Right-Pointing-Pointer Designed to achieve an order of magnitude enhancement over traditional reactor. Black-Right-Pointing-Pointer Al{sub 2}O{sub 3} and hydrotalcite coatings capture and trap pollutants giving additional 30% TOC removal. Black-Right-Pointing-Pointer High surface area coating prevents polarization and improves membrane separation and life.« less

Precise Nuclear Data Measurements Possible with the NIFFTE fissionTPC for Advanced Reactor Designs

NASA Astrophysics Data System (ADS)

Towell, Rusty; Niffte Collaboration

2015-10-01

The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) Collaboration has applied the proven technology of Time Projection Chambers (TPC) to the task of precisely measuring fission cross sections. With the NIFFTE fission TPC, precise measurements have been made during the last year at the Los Alamos Neutron Science Center from both U-235 and Pu-239 targets. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow systematics to be controlled at the level of 1%. The fissionTPC performance will be presented. These results are critical to the development of advanced uranium-fueled reactors. However, there are clear advantages to developing thorium-fueled reactors such as Liquid Fluoride Thorium Reactors over uranium-fueled reactors. These advantages include improved reactor safety, minimizing radioactive waste, improved reactor efficiency, and enhanced proliferation resistance. The potential for using the fissionTPC to measure needed cross sections important to the development of thorium-fueled reactors will also be discussed.

Johnson Noise Thermometry for Advanced Small Modular Reactors

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

Britton Jr, Charles L; Roberts, Michael; Bull, Nora D

Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurementmore » due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.« less

Lasers in Synthesis, Characterization and Processing of Diamond, International Center, Tashkent, Uzbekistan, 6-9 October 1997

DTIC Science & Technology

1997-01-01

Chemistry Division, Code 6174 Materiaux Leninsky prospekt, 53 Gas/Surface Dinamics Section et des Hautes Pressions Moscow 117924, Russia Washington, D.C...reactor for diamond CVD. Strengths and limitations of this and the various alternative H atom detection methods will be summarised, before

NDI (Nondestructive Inspection) Oriented Corrosion Control for Army Aircraft. Phase 1. Inspection Methods

DTIC Science & Technology

1989-07-01

Appendices A and B and are provided as cover sheets from each item rather than completc packages. The Pamplet Series materials were furnished as camera-ready...34Stational Neutron Radiography System for Aircraft Reliability and Maintainability." G. A. Technologies Brochure , Triga Reactor Division, San Diego

Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

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

Tan, Lizhen; Yang, Ying; Sridharan, K.

2015-12-01

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computationalmore » tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe 2+) irradiation.« less

An advanced carbon reactor subsystem for carbon dioxide reduction

NASA Technical Reports Server (NTRS)

Noyes, Gary P.; Cusick, Robert J.

1986-01-01

An evaluation is presented of the development status of an advanced carbon-reactor subsystem (ACRS) for the production of water and dense, solid carbon from CO2 and hydrogen, as required in physiochemical air revitalization systems for long-duration manned space missions. The ACRS consists of a Sabatier Methanation Reactor (SMR) that reduces CO2 with hydrogen to form methane and water, a gas-liquid separator to remove product water from the methane, and a Carbon Formation Reactor (CFR) to pyrolize methane to carbon and hydrogen; the carbon is recycled to the SMR, while the produce carbon is periodically removed from the CFR. A preprototype ACRS under development for the NASA Space Station is described.

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

NASA Technical Reports Server (NTRS)

Bowles, K. J.

1974-01-01

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

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

Block, R.C.; Feiner, F.

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

Advanced Instrumentation for Transient Reactor Testing

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

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

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

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

DOE PAGES

Melin, Alexander M.; Kisner, Roger A.

2018-04-03

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

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

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

Melin, Alexander M.; Kisner, Roger A.

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

A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility

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

S. Khericha

2010-12-01

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

Chemistry Division annual progress report for period ending April 30, 1993

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

Poutsma, M.L.; Ferris, L.M.; Mesmer, R.E.

1993-08-01

The Chemistry Division conducts basic and applied chemical research on projects important to DOE`s missions in sciences, energy technologies, advanced materials, and waste management/environmental restoration; it also conducts complementary research for other sponsors. The research are arranged according to: coal chemistry, aqueous chemistry at high temperatures and pressures, geochemistry, chemistry of advanced inorganic materials, structure and dynamics of advanced polymeric materials, chemistry of transuranium elements and compounds, chemical and structural principles in solvent extraction, surface science related to heterogeneous catalysis, photolytic transformations of hazardous organics, DNA sequencing and mapping, and special topics.

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.

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.

Simulator platform for fast reactor operation and safety technology demonstration

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

Vilim, R. B.; Park, Y. S.; Grandy, C.

2012-07-30

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe responsemore » to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.« less

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2013 CFR

2013-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2012 CFR

2012-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in the...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2014 CFR

2014-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

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

Monteleone, S.

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

Hurricane Research Division of AOML/NOAA

Science.gov Websites

Statement The mission of NOAA's Hurricane Research Division (HRD) is to advance the understanding and Learn More. What's New Links of Interest Hurricane Field Program Current Hurricane Data Hurricane FAQ

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

Multiple access techniques and spectrum utilization of the GLOBALSTAR mobile satellite system

NASA Astrophysics Data System (ADS)

Louie, Ming; Cohen, Michel; Rouffet, Denis; Gilhousen, Klein S.

The GLOBALSTAR System is a Low Earth Orbit (LEO) satellite-based mobile communications system that is interoperable with the current and future Public Land Mobile Network (PLMN). The GLOBALSTAR System concept is based upon technological advancement in two key areas: (1) the advancement in LEO satellite technology; (2) the advancement in cellular telephone technology, including the commercial applications of Code Division Multiple Access (CDMA) technologies, and of the most recent progress in Time Division Multiple Access technologies. The GLOBALSTAR System uses elements of CDMA, Frequency Division Multiple Access (FDMA), and Time Division Multiple Access (TDMA) technology, combining with satellite Multiple Beam Antenna (MBA) technology, to arrive at one of the most efficient modulation and multiple access system ever proposed for a satellite communications system. The technology used in GLOBALSTAR exploits the following techniques in obtaining high spectral efficiency and affordable cost per channel, with minimum coordination among different systems: power control, in open and closed loops, voice activation, spot beam satellite antenna for frequency reuse, weighted satellite antenna gain, multiple satellite coverage, and handoff between satellites. The GLOBALSTAR system design will use the following frequency bands: 1610-1626.5 MHz for up-link and 2483.5-2500 MHz for down-link.

10 CFR 72.186 - Change to physical security and safeguards contingency plans.

Code of Federal Regulations, 2012 CFR

2012-01-01

... approval for a period of three years from the date of the change, and shall, within two months after the change is made, submit a report addressed to Director, Division of Spent Fuel Storage and Transportation... THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED...

10 CFR 72.186 - Change to physical security and safeguards contingency plans.

Code of Federal Regulations, 2014 CFR

2014-01-01

... approval for a period of three years from the date of the change, and shall, within two months after the change is made, submit a report addressed to Director, Division of Spent Fuel Storage and Transportation... THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED...

10 CFR 72.186 - Change to physical security and safeguards contingency plans.

Code of Federal Regulations, 2013 CFR

2013-01-01

... approval for a period of three years from the date of the change, and shall, within two months after the change is made, submit a report addressed to Director, Division of Spent Fuel Storage and Transportation... THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED...

10 CFR 72.186 - Change to physical security and safeguards contingency plans.

Code of Federal Regulations, 2011 CFR

2011-01-01

... approval for a period of three years from the date of the change, and shall, within two months after the change is made, submit a report addressed to Director, Division of Spent Fuel Storage and Transportation... THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED...

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...

75 FR 74105 - Dominion Virginia Power and Old Dominion Electric Cooperative; Notice of Consideration of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-30

... Patel, Project Manager, US-APWR Projects Branch, Division of New Reactor Licensing, Office of New.... Before issuing the amendment, regardless of whether a hearing is requested, the Commission will make a... make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take...

Mass and Performance Estimates for 5 to 1000 kW(e) Nuclear Reactor Power Systems for Space Applications

DTIC Science & Technology

1990-12-01

Albany Street Cambridge, MA 02139 Dave Berwald Grumman Aerospace Corporation MS B20-05 Bethpage, NY 11714 F. Best Assistant Professor Texas A&M... Zielinski U. S. Department of Energy SAN-ACR Division 13333 Broadway Oakland, CA 94612 G. L. Zigler Science & Engineering Associates 6301

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

Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors

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

William Richins; Stephen Novascone; Cheryl O'Brien

Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors William Richins1, Stephen Novascone1, and Cheryl O’Brien1 1Idaho National Laboratory, US Dept. of Energy, Idaho Falls, Idaho, USA, e-mail: William.Richins@inl.gov The Idaho National Laboratory (INL, USA) and IASMiRT sponsored an international forum Nov 5-6, 2008 in Porvoo, Finland for nuclear industry, academic, and regulatory representatives to identify structural issues in current and future advanced reactor design, especially for extreme conditions and external threats. The purpose of this Topical Workshop was to articulate research, engineering, and regulatory Code development needs. The topics addressed by the Workshop were selectedmore » to address critical industry needs specific to advanced reactor structures that have long lead times and can be the subject of future SMiRT technical sessions. The topics were; 1) structural/materials needs for extreme conditions and external threats in contemporary (Gen. III) and future (Gen. IV and NGNP) advanced reactors and 2) calibrating simulation software and methods that address topic 1 The workshop discussions and research needs identified are presented. The Workshop successfully produced interactive discussion on the two topics resulting in a list of research and technology needs. It is recommended that IASMiRT communicate the results of the discussion to industry and researchers to encourage new ideas and projects. In addition, opportunities exist to retrieve research reports and information that currently exists, and encourage more international cooperation and collaboration. It is recommended that IASMiRT continue with an off-year workshop series on select topics.« less

Enhanced In-Pile Instrumentation at the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

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

2012-08-01

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

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

Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

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

Fromm, Bradley; Hauch, Benjamin; Sridharan, Kumar

2016-12-01

A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulationmore » tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.« less

The ORNL Chemical Technology Division, 1950-1994

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

Jolley, R.L.; Genung, R.K.; McNeese, L.E.

1994-10-01

This document attempts to reconstruct the role played by the Chemical Technology Division (Chem Tech) of the Oak Ridge National Laboratory (ORNL) in the atomic era since the 1940`s related to the development and production of nuclear weapons and power reactors. Chem Tech`s early contributions were landmark pioneering studies. Unknown and dimly perceived problems like chemical hazards, radioactivity, and criticality had to be dealt with. New chemical concepts and processes had to be developed to test the new theories being developed by physicists. New engineering concepts had to be developed and demonstrated in order to build facilities and equipment thatmore » had never before been attempted. Chem Tech`s role was chemical separations, especially uranium and plutonium, and nuclear fuel reprocessing. With diversification of national and ORNL missions, Chem Tech undertook R&D studies in many areas including biotechnology; clinical and environmental chemistry; nuclear reactors; safety regulations; effective and safe waste management and disposal; computer modeling and informational databases; isotope production; and environmental control. The changing mission of Chem Tech are encapsulated in the evolving activities.« less

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

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

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

2006-10-01

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

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

78 FR 5840 - Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-28

... University of Illinois Advanced TRIGA Reactor, License No. R-115 The U.S. Nuclear Regulatory Commission (NRC) is noticing the termination of Facility Operating License No. R-115, for the University of Illinois... Operating License No. R-115 is terminated. The above referenced documents may be examined, and/or copied for...

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

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

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

Recent Advances in Pd-Based Membranes for Membrane Reactors.

PubMed

Arratibel Plazaola, Alba; Pacheco Tanaka, David Alfredo; Van Sint Annaland, Martin; Gallucci, Fausto

2017-01-01

Palladium-based membranes for hydrogen separation have been studied by several research groups during the last 40 years. Much effort has been dedicated to improving the hydrogen flux of these membranes employing different alloys, supports, deposition/production techniques, etc. High flux and cheap membranes, yet stable at different operating conditions are required for their exploitation at industrial scale. The integration of membranes in multifunctional reactors (membrane reactors) poses additional demands on the membranes as interactions at different levels between the catalyst and the membrane surface can occur. Particularly, when employing the membranes in fluidized bed reactors, the selective layer should be resistant to or protected against erosion. In this review we will also describe a novel kind of membranes, the pore-filled type membranes prepared by Pacheco Tanaka and coworkers that represent a possible solution to integrate thin selective membranes into membrane reactors while protecting the selective layer. This work is focused on recent advances on metallic supports, materials used as an intermetallic diffusion layer when metallic supports are used and the most recent advances on Pd-based composite membranes. Particular attention is paid to improvements on sulfur resistance of Pd based membranes, resistance to hydrogen embrittlement and stability at high temperature.

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.

Chemical Technology Division: Progress report, January 1, 1987--June 30, 1988

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

Not Available

1989-02-01

This progress report summarizes the research and development efforts conducted in the Chemical Technology Division (Chem Tech) during the period January 1, 1987, to June 30, 1988. The following major areas are covered: waste management and environmental programs, radiochemical and reactor engineering programs, basic science and technology, Nuclear Regulatory Commission programs, and administrative resources and facilities. The Administrative Summary, an appendix, presents a comprehensive listing of publications, oral presentations, awards and recognitions, and patents of Chem Tech staff members during this period. A staffing level and financial summary and lists of seminars and Chem Tech consultants for the period aremore » also included.« less

Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

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

Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi

2013-11-29

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implementmore » a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.« less

Year One Summary of X-energy Pebble Fuel Development at ORNL

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

Helmreich, Grant W.; Hunn, John D.; McMurray, Jake W.

2017-06-01

The Advanced Reactor Concepts X-energy (ARC-Xe) Pebble Fuel Development project at Oak Ridge National Laboratory (ORNL) has successfully completed its first year, having made excellent progress in accomplishing programmatic objectives. The primary focus of research at ORNL in support of X-energy has been the training of X-energy fuel fabrication engineers and the establishment of US pebble fuel production capabilities able to supply the Xe-100 pebble-bed reactor. These efforts have been strongly supported by particle fuel fabrication and characterization expertise present at ORNL from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program.

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

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

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

1998-09-01

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

NASA. Lewis Research Center Advanced Modulation and Coding Project: Introduction and overview

NASA Technical Reports Server (NTRS)

Budinger, James M.

1992-01-01

The Advanced Modulation and Coding Project at LeRC is sponsored by the Office of Space Science and Applications, Communications Division, Code EC, at NASA Headquarters and conducted by the Digital Systems Technology Branch of the Space Electronics Division. Advanced Modulation and Coding is one of three focused technology development projects within the branch's overall Processing and Switching Program. The program consists of industry contracts for developing proof-of-concept (POC) and demonstration model hardware, university grants for analyzing advanced techniques, and in-house integration and testing of performance verification and systems evaluation. The Advanced Modulation and Coding Project is broken into five elements: (1) bandwidth- and power-efficient modems; (2) high-speed codecs; (3) digital modems; (4) multichannel demodulators; and (5) very high-data-rate modems. At least one contract and one grant were awarded for each element.

Overview of NASA Glenn Research Center's Communications and Intelligent Systems Division

NASA Technical Reports Server (NTRS)

Miranda, Felix A.

2016-01-01

The Communications and Intelligent Systems Division provides expertise, plans, conducts and directs research and engineering development in the competency fields of advanced communications and intelligent systems technologies for application in current and future aeronautics and space systems.

78 FR 54255 - HRSA's Bureau of Health Professions Advanced Education Nursing Traineeship Program

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-03

... of Health Professions Advanced Education Nursing Traineeship Program AGENCY: Health Resources and... announcing a change to its Advanced Education Nursing Traineeship (AENT) program. Effective fiscal year (FY... Wasserman, DrPH, RN, Advanced Nursing Education Branch Chief, Division of Nursing, Bureau of Health...

Multichannel demultiplexer/demodulator technologies for future satellite communication systems

NASA Technical Reports Server (NTRS)

Ivancic, William D.; Budinger, James M.; Staples, Edward J.; Abramovitz, Irwin; Courtois, Hector A.

1992-01-01

NASA-Lewis' Space Electronics Div. supports ongoing research in advanced satellite communication architectures, onboard processing, and technology development. Recent studies indicate that meshed VSAT (very small aperture terminal) satellite communication networks using FDMA (frequency division multiple access) uplinks and TDMA (time division multiplexed) downlinks are required to meet future communication needs. One of the critical advancements in such a satellite communication network is the multichannel demultiplexer/demodulator (MCDD). The progress is described which was made in MCDD development using either acousto-optical, optical, or digital technologies.

48 CFR 2032.402 - General.

Code of Federal Regulations, 2010 CFR

2010-10-01

... Chief Financial Officer, Division of Accounting and Finance, to ensure completeness of contractor... REQUIREMENTS CONTRACT FINANCING Advance Payments for Non-Commercial Items 2032.402 General. (a) The contracting... contract terms concerning advance payments. (b) Before authorizing any advance payment agreements, except...

Health physics aspects of advanced reactor licensing reviews

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

Hinson, C.S.

1995-03-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovativemore » design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.« less

New reactor technology: safety improvements in nuclear power systems.

PubMed

Corradini, M L

2007-11-01

Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-02

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... of pressurized water reactor spent nuclear fuel (SNF) in transportation packages and storage casks... for the licensing basis, (b) provide recommendations regarding advanced isotopic depletion and...

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

After Action Report: Advanced Test Reactor Complex 2015 Evaluated Drill October 6, 2015

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

Holmes, Forest Howard

2015-11-01

The Advanced Test Reactor (ATR) Complex, operated by Battelle Energy Alliance, LLC, at the Idaho National Laboratory (INL) conducted an evaluated drill on October 6, 2015, to allow the ATR Complex emergency response organization (ERO) to demonstrate the ability to respond to and mitigate an emergency by implementing the requirements of DOE O 151.1C, “Comprehensive Emergency Management System.”

ADVANCED HEAT TRANSFER TEST FACILITY, TRA666A. ELEVATIONS. ROOF FRAMING PLAN. ...

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

ADVANCED HEAT TRANSFER TEST FACILITY, TRA-666A. ELEVATIONS. ROOF FRAMING PLAN. CONCRETE BLOCK SIDING. SLOPED ROOF. ROLL-UP DOOR. AIR INTAKE ENCLOSURE ON NORTH SIDE. F.C. TORKELSON 842-MTR-666-A5, 8/1966. INL INDEX NO. 531-0666-00-851-152258, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

An Update on Improvements to NiCE Support for PROTEUS

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

Bennett, Andrew; McCaskey, Alexander J.; Billings, Jay Jay

2015-09-01

The Department of Energy Office of Nuclear Energy's Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has supported the development of the NEAMS Integrated Computational Environment (NiCE), a modeling and simulation workflow environment that provides services and plugins to facilitate tasks such as code execution, model input construction, visualization, and data analysis. This report details the development of workflows for the reactor core neutronics application, PROTEUS. This advanced neutronics application (primarily developed at Argonne National Laboratory) aims to improve nuclear reactor design and analysis by providing an extensible and massively parallel, finite-element solver for current and advanced reactor fuel neutronicsmore » modeling. The integration of PROTEUS-specific tools into NiCE is intended to make the advanced capabilities that PROTEUS provides more accessible to the nuclear energy research and development community. This report will detail the work done to improve existing PROTEUS workflow support in NiCE. We will demonstrate and discuss these improvements, including the development of flexible IO services, an improved interface for input generation, and the addition of advanced Fortran development tools natively in the platform.« less

RECENT ADVANCES IN THE MODELING OF AIRBORNE SUBSTANCES

EPA Science Inventory

Since the 1950's, the primary mission of the Atmospheric Modeling Division has been to develop and evaluate air quality simulation models. While the Division has traditionally focused the research on the meteorological aspects of these models, this focus has expanded in recent...

Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

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

Tome, Carlos N; Caro, J A; Lebensohn, R A

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating themore » phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.« less

Advanced ceramic materials for next-generation nuclear applications

NASA Astrophysics Data System (ADS)

Marra, John

2011-10-01

The nuclear industry is at the eye of a 'perfect storm' with fuel oil and natural gas prices near record highs, worldwide energy demands increasing at an alarming rate, and increased concerns about greenhouse gas (GHG) emissions that have caused many to look negatively at long-term use of fossil fuels. This convergence of factors has led to a growing interest in revitalization of the nuclear power industry within the United States and across the globe. Many are surprised to learn that nuclear power provides approximately 20% of the electrical power in the US and approximately 16% of the world-wide electric power. With the above factors in mind, world-wide over 130 new reactor projects are being considered with approximately 25 new permit applications in the US. Materials have long played a very important role in the nuclear industry with applications throughout the entire fuel cycle; from fuel fabrication to waste stabilization. As the international community begins to look at advanced reactor systems and fuel cycles that minimize waste and increase proliferation resistance, materials will play an even larger role. Many of the advanced reactor concepts being evaluated operate at high-temperature requiring the use of durable, heat-resistant materials. Advanced metallic and ceramic fuels are being investigated for a variety of Generation IV reactor concepts. These include the traditional TRISO-coated particles, advanced alloy fuels for 'deep-burn' applications, as well as advanced inert-matrix fuels. In order to minimize wastes and legacy materials, a number of fuel reprocessing operations are being investigated. Advanced materials continue to provide a vital contribution in 'closing the fuel cycle' by stabilization of associated low-level and high-level wastes in highly durable cements, ceramics, and glasses. Beyond this fission energy application, fusion energy will demand advanced materials capable of withstanding the extreme environments of high-temperature plasma systems. Fusion reactors will likely depend on lithium-based ceramics to produce tritium that fuels the fusion plasma, while high-temperature alloys or ceramics will contain and control the hot plasma. All the while, alloys, ceramics, and ceramic-related processes continue to find applications in the management of wastes and byproducts produced by these processes.

Creating Age Asymmetry: Consequences of Inheriting Damaged Goods in Mammalian Cells.

PubMed

Moore, Darcie L; Jessberger, Sebastian

2017-01-01

Accumulating evidence suggests that mammalian cells asymmetrically segregate cellular components ranging from genomic DNA to organelles and damaged proteins during cell division. Asymmetric inheritance upon mammalian cell division may be specifically important to ensure cellular fitness and propagate cellular potency to individual progeny, for example in the context of somatic stem cell division. We review here recent advances in the field and discuss potential effects and underlying mechanisms that mediate asymmetric segregation of cellular components during mammalian cell division. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

Tan, Lizhen; Yang, Ying; Tyburska-Puschel, Beata

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools)more » is an important path to more efficient alloy development and process optimization. Ferritic-martensitic (FM) steels are important structural materials for nuclear reactors due to their advantages over other applicable materials like austenitic stainless steels, notably their resistance to void swelling, low thermal expansion coefficients, and higher thermal conductivity. However, traditional FM steels exhibit a noticeable yield strength reduction at elevated temperatures above ~500°C, which limits their applications in advanced nuclear reactors which target operating temperatures at 650°C or higher. Although oxide-dispersion-strengthened (ODS) ferritic steels have shown excellent high-temperature performance, their extremely high cost, limited size and fabricability of products, as well as the great difficulty with welding and joining, have limited or precluded their commercial applications. Zirconium has shown many benefits to Fe-base alloys such as grain refinement, improved phase stability, and reduced radiation-induced segregation. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of a new generation of Zr-bearing ferritic alloys to be fabricated using conventional steelmaking practices, which have excellent radiation resistance and enhanced high-temperature creep performance greater than Grade 91.« less

Characterization of contaminant removal by an optical strip material

NASA Astrophysics Data System (ADS)

Hamilton, James P.; Frigo, S. P.; Caroll, Brenden J.; Assoufidyen, L.; Lewis, Matthew S.; Cook, Russell E.; de Carlo, F.

2001-03-01

Department of Chemistry and Engineering Physics, University of Wisconsin-Platteville, Platteville, WI 53818 Advanced Photon Source, X-Ray Facilities Division, Argonne National Laboratory, Advanced Photon Source, User Program Division, Argonne National Laboratory, *Electron Microscopy Center, Materials Science Division, Argonne National Laboratory, Argonne National Laboratory, 9700 S. Cass Ave., Argonne IL 60439-4856 USA A novel optical strip coating material, Opticlean, has been shown to safely remove fingerprints, particles and contamination from a variety of optical surfaces including coated glass, Si and first surface mirrors. Contaminant removal was monitored by Nomarski, Atomic Force and Scanning Electron Microscopy. Sub-micron features on diffraction gratings and silicon wafers were also cleaned without leaving light scattering particles on the surface. **This work was supported in part by the U.S. Department of Energy, Basic Energy Sciences-Materials Sciences, under contract no. W-31-109-ENG-38. The authors acknowledge the support and facilities provided by the Advanced Photon Source and the Electron Microscopy Center at Argonne National Laboratory.

A thermodynamic approach for advanced fuels of gas-cooled reactors

NASA Astrophysics Data System (ADS)

Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

2005-09-01

For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

Solid State Division progress report for period ending September 30, 1993

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

Green, P.H.; Hinton, L.W.

1994-08-01

This report covers research progress in the Solid State Division from April 1, 1992, to September 30, 1993. During this period, the division conducted a broad, interdisciplinary materials research program with emphasis on theoretical solid state physics, neutron scattering, synthesis and characterization of materials, ion beam and laser processing, and the structure of solids and surfaces. This research effort was enhanced by new capabilities in atomic-scale materials characterization, new emphasis on the synthesis and processing of materials, and increased partnering with industry and universities. The theoretical effort included a broad range of analytical studies, as well as a new emphasismore » on numerical simulation stimulated by advances in high-performance computing and by strong interest in related division experimental programs. Superconductivity research continued to advance on a broad front from fundamental mechanisms of high-temperature superconductivity to the development of new materials and processing techniques. The Neutron Scattering Program was characterized by a strong scientific user program and growing diversity represented by new initiatives in complex fluids and residual stress. The national emphasis on materials synthesis and processing was mirrored in division research programs in thin-film processing, surface modification, and crystal growth. Research on advanced processing techniques such as laser ablation, ion implantation, and plasma processing was complemented by strong programs in the characterization of materials and surfaces including ultrahigh resolution scanning transmission electron microscopy, atomic-resolution chemical analysis, synchrotron x-ray research, and scanning tunneling microscopy.« less

138. ARAII Building ARA606 floor plan for remodel as Inel ...

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

138. ARA-II Building ARA-606 floor plan for remodel as Inel Welding Laboratory. Shows room divisions and welding stations to be installed. Aerojet Nuclear Company 1375-ARA-II-606-E-2. Date: June 1976. Ineel index code no. 070-0606-10-400-156552. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Development of Advanced 9Cr Ferritic-Martensitic Steels and Austenitic Stainless Steels for Sodium-Cooled Fast Reactor

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

Sham, Sam; Tan, Lizhen; Yamamoto, Yukinori

2013-01-01

Ferritic-martensitic (FM) steel Grade 92, with or without thermomechanical treatment (TMT), and austenitic stainless steels HT-UPS (high-temperature ultrafine precipitate strengthening) and NF709 were selected as potential candidate structural materials in the U.S. Sodium-cooled Fast Reactor (SFR) program. The objective is to develop advanced steels with improved properties as compared with reference materials such as Grade 91 and Type 316H steels that are currently in nuclear design codes. Composition modification and/or processing optimization (e.g., TMT and cold-work) were performed to improve properties such as resistance to thermal aging, creep, creep-fatigue, fracture, and sodium corrosion. Testings to characterize these properties for themore » advanced steels were conducted by the Idaho National Laboratory, the Argonne National Laboratory and the Oak Ridge National Laboratory under the U.S. SFR program. This paper focuses on the resistance to thermal aging and creep of the advanced steels. The advanced steels exhibited up to two orders of magnitude increase in creep life compared to the reference materials. Preliminary results on the weldment performance of the advanced steels are also presented. The superior performance of the advanced steels would improve reactor design flexibility, safety margins and economics.« less

The Advanced Software Development and Commercialization Project

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

Gallopoulos, E.; Canfield, T.R.; Minkoff, M.

1990-09-01

This is the first of a series of reports pertaining to progress in the Advanced Software Development and Commercialization Project, a joint collaborative effort between the Center for Supercomputing Research and Development of the University of Illinois and the Computing and Telecommunications Division of Argonne National Laboratory. The purpose of this work is to apply techniques of parallel computing that were pioneered by University of Illinois researchers to mature computational fluid dynamics (CFD) and structural dynamics (SD) computer codes developed at Argonne. The collaboration in this project will bring this unique combination of expertise to bear, for the first time,more » on industrially important problems. By so doing, it will expose the strengths and weaknesses of existing techniques for parallelizing programs and will identify those problems that need to be solved in order to enable wide spread production use of parallel computers. Secondly, the increased efficiency of the CFD and SD codes themselves will enable the simulation of larger, more accurate engineering models that involve fluid and structural dynamics. In order to realize the above two goals, we are considering two production codes that have been developed at ANL and are widely used by both industry and Universities. These are COMMIX and WHAMS-3D. The first is a computational fluid dynamics code that is used for both nuclear reactor design and safety and as a design tool for the casting industry. The second is a three-dimensional structural dynamics code used in nuclear reactor safety as well as crashworthiness studies. These codes are currently available for both sequential and vector computers only. Our main goal is to port and optimize these two codes on shared memory multiprocessors. In so doing, we shall establish a process that can be followed in optimizing other sequential or vector engineering codes for parallel processors.« less

Proceedings of a Symposium on Advanced Compact Reactor Systems

NASA Technical Reports Server (NTRS)

1983-01-01

Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

Radio-toxicity of spent fuel of the advanced heavy water reactor.

PubMed

Anand, S; Singh, K D S; Sharma, V K

2010-01-01

The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR.

RESEARCH AND TECHNOLOGY DIVISION REPORT FOR 1966.

ERIC Educational Resources Information Center

BAUM, C.

THE WORK OF THE RESEARCH AND TECHNOLOGY DIVISION OF SYSTEM DEVELOPMENT CORPORATION DURING 1966 IS REPORTED. THE PROGRESS OF VARIOUS STUDIES AND ACTIVITIES DISCUSSED IN THE REPORT WERE ADVANCED PROGRAMING, INFORMATION PROCESSING RESEARCH, PROGRAMING SYSTEMS, DATA BASE SYSTEMS. LANGUAGE PROCESSING AND RETRIEVAL, BEHAVIORAL GAMING AND SIMULATION…

78 FR 48915 - Sunshine Act Meetings Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-12

... app; update on the revised ``STEM Education Data and Trends'' online tool Discussion of the Science...), Division of Astronomical Sciences (AST): Revised Baseline for the Construction of the Advanced Technology...), Division of Astronomical Sciences (AST): Increase in Pass-through Authority for Construction of the Atacama...

10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

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

Bohachek, Randolph Charles

2015-09-01

The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactorsmore » is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.« less

10 CFR 73.72 - Requirement for advance notice of shipment of formula quantities of strategic special nuclear...

Code of Federal Regulations, 2010 CFR

2010-01-01

..., or irradiated reactor fuel. 73.72 Section 73.72 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED... strategic significance, or irradiated reactor fuel. (a) A licensee, other than one specified in paragraph (b... strategic significance, or irradiated reactor fuel required to be protected in accordance with § 73.37...

10 CFR 73.72 - Requirement for advance notice of shipment of formula quantities of strategic special nuclear...

Code of Federal Regulations, 2011 CFR

2011-01-01

..., or irradiated reactor fuel. 73.72 Section 73.72 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED... strategic significance, or irradiated reactor fuel. (a) A licensee, other than one specified in paragraph (b... strategic significance, or irradiated reactor fuel required to be protected in accordance with § 73.37...

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

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

Davenport, Michael; Petti, D. A.; Palmer, Joe

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experimentsmore » are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.« less

Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

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

Carbajo, Juan J; Qualls, A L

2016-01-01

INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a smallmore » version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.« 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.

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

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

O'Brien, James E.; Sabharwall, Piyush; Yoon, Su -Jong

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs)more » at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation needs. The experimental database will guide development of appropriate predictive methods and be available for code verification and validation (V&V) related to these systems.« less

Effects of Levels of Automation for Advanced Small Modular Reactors: Impacts on Performance, Workload, and Situation Awareness

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

Johanna Oxstrand; Katya Le Blanc

The Human-Automation Collaboration (HAC) research effort is a part of the Department of Energy (DOE) sponsored Advanced Small Modular Reactor (AdvSMR) program conducted at Idaho National Laboratory (INL). The DOE AdvSMR program focuses on plant design and management, reduction of capital costs as well as plant operations and maintenance costs (O&M), and factory production costs benefits.

2014 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

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

Lewis, Mike

2015-02-01

This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

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

Advanced In-Pile Instrumentation for Materials Testing Reactors

NASA Astrophysics Data System (ADS)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T. C.; Chase, B. M.; Davis, K. L.; Palmer, A. J.; Schley, R. S.

2014-08-01

The U.S. Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified, and the progress of other development efforts is summarized. As reported in this paper, INL researchers are currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating `advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers, are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors.

Radiogenic lead as coolant, reflector and moderator in advanced fast reactors

NASA Astrophysics Data System (ADS)

Kulikov, E. G.

2017-01-01

Main purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors. When performing the study, thermal, physical and neutron-physical properties of natural and radiogenic lead were analyzed. The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and accelerator-driven systems (ADS) makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket, which enables effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.

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

NASA Astrophysics Data System (ADS)

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

1992-03-01

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

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

Instrumentation and Controls Division Progress Report for the Period of July 1, 1994 to December 31, 1997: Publications, Presentations, Activities, and Awards

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

McDonald, D.W.

This report contains a record of publishing and other activities in the Oak Ridge National Laboratory (ORNL) Instrumentation and Controls (I&C) Division for the period of July 1, 1994, to December31, 1997. It is a companion volume to Working Together on New Horizons: Instrumentation and Controls Division Progress Report for the Period of July 1, 1994, to December 31, 1997 (OR.NLA4-6530). Working Together on New Horizons contains illustrated summaries of some of the projects under way in I&C Division. Both books can be obtained by contacting C. R. Brittain (brittain@ornl. gov), P.O. Box 2008, Oak Ridge, TN 37831-6005. l&C Divisionmore » Mission and Vision I&C Division develops and maintains techniques, instruments, and systems that lead to a better understanding of nature and harnessing of natural phenomena for the benefit of humankind. We have dedicated ourselves to accelerating the advancement of science and the transfer of those advancements into products and processes that benefit U.S. industry and enhance the security of our citizens.« less

Phenomena Important in Molten Salt Reactor Simulations

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

Diamond, David J.; Brown, Nicholas R.; Denning, Richard

The U.S. Nuclear Regulatory Commission (NRC) is preparing for the future licensing of advanced reactors that will be very different from current light water reactors. Part of the NRC preparation strategy is to identify the simulation tools that will be used for confirmatory safety analysis of normal operation and abnormal situations in those reactors. This report advances that strategy for reactors that will use molten salts (MSRs). This includes reactors with the fuel within the salt as well as reactors using solid fuel. Although both types are discussed in this report, the emphasis is on those reactors with liquid fuelmore » because of the perception that solid-fuel MSRs will be significantly easier to simulate. These liquid-fuel reactors include thermal and fast neutron spectrum alternatives. The specific designs discussed in the report are a subset of many designs being considered in the U.S. and elsewhere but they are considered the most likely to submit information to the NRC in the near future. The objective herein, is to understand the design of proposed molten salt reactors, how they will operate under normal or transient/accident conditions, and what will be the corresponding modeling needs of simulation tools that consider neutronics, heat transfer, fluid dynamics, and material composition changes in the molten salt. These tools will enable the NRC to eventually carry out confirmatory analyses that examine the validity and accuracy of applicant’s calculations and help determine the margin of safety in plant design.« less

Analytical Chemistry Division annual progress report for period ending December 31, 1988

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

Not Available

The Analytical Chemistry Division of Oak Ridge National Laboratory (ORNL) is a large and diversified organization. As such, it serves a multitude of functions for a clientele that exists both in and outside of ORNL. These functions fall into the following general categories: (1) Analytical Research, Development, and Implementation. The division maintains a program to conceptualize, investigate, develop, assess, improve, and implement advanced technology for chemical and physicochemical measurements. Emphasis is on problems and needs identified with ORNL and Department of Energy (DOE) programs; however, attention is also given to advancing the analytical sciences themselves. (2) Programmatic Research, Development, andmore » Utilization. The division carries out a wide variety of chemical work that typically involves analytical research and/or development plus the utilization of analytical capabilities to expedite programmatic interests. (3) Technical Support. The division performs chemical and physicochemical analyses of virtually all types. The Analytical Chemistry Division is organized into four major sections, each of which may carry out any of the three types of work mentioned above. Chapters 1 through 4 of this report highlight progress within the four sections during the period January 1 to December 31, 1988. A brief discussion of the division's role in an especially important environmental program is given in Chapter 5. Information about quality assurance, safety, and training programs is presented in Chapter 6, along with a tabulation of analyses rendered. Publications, oral presentations, professional activities, educational programs, and seminars are cited in Chapters 7 and 8.« less

Cell division cycle 45 promotes papillary thyroid cancer progression via regulating cell cycle.

PubMed

Sun, Jing; Shi, Run; Zhao, Sha; Li, Xiaona; Lu, Shan; Bu, Hemei; Ma, Xianghua

2017-05-01

Cell division cycle 45 was reported to be overexpressed in some cancer-derived cell lines and was predicted to be a candidate oncogene in cervical cancer. However, the clinical and biological significance of cell division cycle 45 in papillary thyroid cancer has never been investigated. We determined the expression level and clinical significance of cell division cycle 45 using The Cancer Genome Atlas, quantitative real-time polymerase chain reaction, and immunohistochemistry. A great upregulation of cell division cycle 45 was observed in papillary thyroid cancer tissues compared with adjacent normal tissues. Furthermore, overexpression of cell division cycle 45 positively correlates with more advanced clinical characteristics. Silence of cell division cycle 45 suppressed proliferation of papillary thyroid cancer cells via G1-phase arrest and inducing apoptosis. The oncogenic activity of cell division cycle 45 was also confirmed in vivo. In conclusion, cell division cycle 45 may serve as a novel biomarker and a potential therapeutic target for papillary thyroid cancer.

Advancing research and practice: the revised APA Division 30 definition of hypnosis.

PubMed

Elkins, Gary R; Barabasz, Arreed F; Council, James R; Spiegel, David

2015-01-01

This article describes the history, rationale, and guidelines for developing a new definition of hypnosis by the Society of Psychological Hypnosis, Division 30 of the American Psychological Association. The definition was developed with the aim of being concise, heuristic, and allowing for alternative theories of the mechanisms (to be determined in empirical scientific study). The definition of hypnosis is presented as well as definitions of the following related terms: hypnotic induction, hypnotizability, and hypnotherapy. The implications for advancing research and practice are discussed. The definitions are presented within the article.

Advancing Research and Practice: The Revised APA Division 30 Definition of Hypnosis.

PubMed

Elkins, Gary R; Barabasz, Arreed F; Council, James R; Spiegel, David

2015-04-01

This article describes the history, rationale, and guidelines for developing a new definition of hypnosis by the Society of Psychological Hypnosis, Division 30 of the American Psychological Association. The definition was developed with the aim of being concise, being heuristic, and allowing for alternative theories of the mechanisms (to be determined in empirical scientific study). The definition of hypnosis is presented as well as definitions of the following related terms: hypnotic induction, hypnotizability, and hypnotherapy. The implications for advancing research and practice are discussed. The definitions are presented within the article.

Advanced Reactor Technologies - Regulatory Technology Development Plan (RTDP)

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

Moe, Wayne L.

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory importance of key DOE reactor research initiatives should be assessed early in the technology development process. Quality assurance requirements supportive of later licensing activities must also be attached to important research activities to ensure resulting data is usable in that context. Early regulatory analysis and licensing approach planning thus provides a significant benefit to the formulation of research plans and also enables the planning and development of a compatible AdvSMR licensing framework, should significant modification be required.« less

Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

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

Moe, Wayne Leland

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory importance of key DOE reactor research initiatives should be assessed early in the technology development process. Quality assurance requirements supportive of later licensing activities must also be attached to important research activities to ensure resulting data is usable in that context. Early regulatory analysis and licensing approach planning thus provides a significant benefit to the formulation of research plans and also enables the planning and development of a compatible AdvSMR licensing framework, should significant modification be required.« less

Apollo - An advanced fuel fusion power reactor for the 21st century

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

Kulcinski, G.L.; Emmert, G.A.; Blanchard, J.P.

1989-03-01

A preconceptual design of a tokamak reactor fueled by a D-He-3 plasma is presented. A low aspect ratio (A=2-4) device is studied here but high aspect ratio devices (A > 6) may also be quite attractive. The Apollo D-He-3 tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The overall efficiency ranges from 37 to 52% depending on whether the bremsstrahlung energy is utilized. The low neutron wall loading (0.1 MW/m/sup 2/) allows a permanent first wall to be designed and the low nuclear decay heat enablesmore » the reactor to be classed as inherently safe. The cost of electricity from Apollo is > 40% lower than electricity from a similar sized DT reactor.« less

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

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

Bucknor, Matthew D.; Grabaskas, David; Brunett, Acacia J.

2016-01-01

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologiesmore » for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Centering on an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive reactor cavity cooling system following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. While this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability for the reactor cavity cooling system (and the reactor system in general) to the postulated transient event.« less

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

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

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

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

DOE PAGES

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; ...

2017-01-24

We report that many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has beenmore » examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Lastly, although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.« less

NRC ARDC Guidance Support Status Report

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

Holbrook, Mark R.

This report provides a summary that reflects the progress and status of proposed regulatory design criteria for advanced non-light water reactor (LWR) designs in accordance with the Level 3 milestone M3AT-17IN2001013 in work package AT-17IN200101. These criteria have been designated as advanced reactor design criteria (ARDC) and they provide guidance to future applicants for addressing the general design criteria (GDC) that are currently applied specifically to LWR designs. This report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of ARDC regulatory guidance for sodium fast reactor (SFR) andmore » modular high-temperature gas-cooled reactor (HTGR) designs. Status Report Organization: Section 2 discusses the origin of the GDC and their application to LWRs. Section 3 addresses the objective of this initiative and how it benefits the advanced non-LWR reactor vendors. Section 4 discusses the scope and structure of the initiative. Section 5 provides background on the U.S. Department of Energy (DOE) ARDC team’s original development of the proposed ARDC that were submitted to the NRC for consideration. Section 6 provides a summary of recent ARDC Phase 2 activities. Appendices A through E document the DOE ARDC team’s public comments on various sections of the NRC’s draft regulatory guide DG–1330, “Guidance for Developing Principal Design Criteria for Non-Light Water Reactors.”« less

Advanced Thermal Batteries.

DTIC Science & Technology

1981-06-01

ADVANCED THERMAL BATTERIES NATIONAL UNION ELECTRIC CORPORATION ADVANCE SCIENCE DIVISION 1201 E. BELL STREET BLXXMINGTON, ILLINOIS 61701 JUNE 1981...December 1978 in: " Advanced Thermal Batteries " AFAPL-TR-78-114 Air Force Aero Propulsion Laboratory Air Force Wright Aeronautical Laboratories Air Force...March 1980 in: " Advanced Thermal Batteries " AFAPL-TR-80-2017 Air Force Aero Propulsion Laboratory Air Force Wright Aeronautical Laboratories Air Force

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

2015 Annual Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Ponds

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

Lewis, Michael George

This report describes conditions and information, as required by the state of Idaho, Department of Environmental Quality Reuse Permit I-161-02, for the Advanced Test Reactor Complex Cold Waste Ponds located at Idaho National Laboratory from November 1, 2014–October 31, 2015. The effective date of Reuse Permit I-161-02 is November 20, 2014 with an expiration date of November 19, 2019.

Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

DOE PAGES

Bartel, N.; Chen, M.; Utgikar, V. P.; ...

2015-04-04

A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimummore » combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.« less

Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

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

Bartel, N.; Chen, M.; Utgikar, V. P.

A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimummore » combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.« less

75 FR 41551 - Meetings of Humanities Panel

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-16

... Psychology in Fellowships, submitted to the Division of Research Programs at the May 4, 2010 deadline. 21... applications for Advanced Social Science Research on Japan in Fellowships, submitted to the Division of...:30 a.m. to 5 p.m. Room: 415. Program: This meeting will review applications for Social Sciences and...

75 FR 14643 - Office of New Reactors; Proposed Standard Review Plan, Branch Technical Position 7-19 on Guidance...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-26

... electronic form, will be posted on the NRC Web site and on the Federal rulemaking Web site http://www... that they do not want publicly disclosed. Federal Rulemaking Web site: Go to http://www.regulations.gov... CONTACT: Mr. Ian C. Jung, Chief, Instrumentation, Controls and Electrical Engineering Branch 2, Division...

75 FR 23820 - Notice of Docketing of Amendment Request for Materials License No. SNM-2506; Northern States...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-04

... INFORMATION CONTACT: Pamela Longmire, Ph.D., Project Manager, Licensing Branch, Division of Spent Fuel Storage... Generating Plant (PINGP), Unit Nos. 1 and 2, site in Goodhue County, Minnesota. The TN-40 cask is currently..., higher burnup spent fuel used in the PINGP reactor as well as associated changes to the ISFSI's technical...

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

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

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

Operational numerical weather prediction on the CYBER 205 at the National Meteorological Center

NASA Technical Reports Server (NTRS)

Deaven, D.

1984-01-01

The Development Division of the National Meteorological Center (NMC), having the responsibility of maintaining and developing the numerical weather forecasting systems of the center, is discussed. Because of the mission of NMC data products must be produced reliably and on time twice daily free of surprises for forecasters. Personnel of Development Division are in a rather unique situation. They must develop new advanced techniques for numerical analysis and prediction utilizing current state-of-the-art techniques, and implement them in an operational fashion without damaging the operations of the center. With the computational speeds and resources now available from the CYBER 205, Development Division Personnel will be able to introduce advanced analysis and prediction techniques into the operational job suite without disrupting the daily schedule. The capabilities of the CYBER 205 are discussed.

Summary of the Advanced Reactor Design Criteria (ARDC) Phase 2 Activities

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

Holbrook, Mark Raymond

This report provides an end-of-year summary reflecting the progress and status of proposed regulatory design criteria for advanced non-LWR designs in accordance with the Level 3 milestone in M3AT-15IN2001017 in work package AT-15IN200101. These criteria have been designated as ARDC, and they provide guidance to future applicants for addressing the GDC that are currently applied specifically to LWR designs. The report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of example adaptations of ARDC for Sodium Fast Reactor (SFR) and modular High Temperature Gas-cooled Reactor (HTGR) designs.

Advanced gray rod control assembly

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

Drudy, Keith J; Carlson, William R; Conner, Michael E

An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber tomore » enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.« less

Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

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

Bernard, J.A.

1989-09-01

This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power onmore » spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.« less

The investigation of a tuberculosis outbreak in the closed environment of a U.S. Navy ship, 1987.

PubMed

DiStasio, A J; Trump, D H

1990-08-01

A sailor on a U.S. Navy ship had smear-positive, cavitary, pulmonary tuberculosis. Contact investigation of the entire ship's crew found 216 new reactors to tuberculin skin test (24.5%) among 881 previously tuberculin-negative sailors. The risk for new infection was highest among sailors in the patient's department (relative risk, 4.4; 95% confidence interval 3.7, 5.3); 95% (15/16) of sailors in his division were new reactors. While crewmembers in all departments were at risk for a new tuberculosis infection, working and berthing in compartments that were distant from those of the index case were protective. The ship's closed ventilation system contributed to the outbreak.

Division Overview

NASA Technical Reports Server (NTRS)

Emerson, Dawn

2016-01-01

This presentation provides an overview of the research and engineering in the competency fields of advanced communications and intelligent systems with emphasis on advanced technologies, architecture definition and system development for application in current and future aeronautics and space systems.

Technological Education for the Rural Community (TERC) Project: Technical Mathematics for the Advanced Manufacturing Technician

ERIC Educational Resources Information Center

McCormack, Sherry L.; Zieman, Stuart

2017-01-01

Hopkinsville Community College's Technological Education for the Rural Community (TERC) project is funded through the National Science Foundation Advanced Technological Education (NSF ATE) division. It is advancing innovative educational pathways for technological education promoted at the community college level serving rural communities to fill…

75 FR 15443 - Advancing the Development of Diagnostic Tests and Biomarkers for Tuberculosis; Public Workshop...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-29

...] Advancing the Development of Diagnostic Tests and Biomarkers for Tuberculosis; Public Workshop; Request for... workshop entitled ``Advancing the Development of Diagnostic Tests and Biomarkers for Tuberculosis (TB... Tuberculosis in the United States, Committee on the Elimination of Tuberculosis in the United States, Division...

Prokaryotic diversity and dynamics in a full-scale municipal solid waste anaerobic reactor from start-up to steady-state conditions.

PubMed

Cardinali-Rezende, Juliana; Colturato, Luís F D B; Colturato, Thiago D B; Chartone-Souza, Edmar; Nascimento, Andréa M A; Sanz, José L

2012-09-01

The prokaryotic diversity of an anaerobic reactor for the treatment of municipal solid waste was investigated over the course of 2 years with the use of 16S rDNA-targeted molecular approaches. The fermentative Bacteroidetes and Firmicutes predominated, and Proteobacteria, Actinobacteria, Tenericutes and the candidate division WWE1 were also identified. Methane production was dominated by the hydrogenotrophic Methanomicrobiales (Methanoculleus sp.) and their syntrophic association with acetate-utilizing and propionate-oxidizing bacteria. qPCR demonstrated the predominance of the hydrogenotrophic over aceticlastic Methanosarcinaceae (Methanosarcina sp. and Methanimicrococcus sp.), and Methanosaetaceae (Methanosaeta sp.) were measured in low numbers in the reactor. According to the FISH and CARD-FISH analyses, Bacteria and Archaea accounted for 85% and 15% of the cells, respectively. Different cell counts for these domains were obtained by qPCR versus FISH analyses. The use of several molecular tools increases our knowledge of the prokaryotic community dynamics from start-up to steady-state conditions in a full-scale MSW reactor. Copyright © 2012 Elsevier Ltd. All rights reserved.

Closure of the R Reactor Disassembly Basin at the SRS

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

Austin, W.E.

The Facilities Disposition Division (FDD) at the Savannah River Site is engaged in planning the deactivation/closure of three of the site's five reactor disassembly basins. Activities are currently underway at R-Reactor Disassembly Basin and will continue with the P and C disassembly basins. The basins still contain the cooling and shielding water that was present when operations ceased. Low concentrations of radionuclides are present, with tritium, Cs-137, and Sr-90 being the major contributors. Although there is no evidence that any of the basins have leaked, the 50-year-old facilities will eventually contaminate the surrounding groundwaters. The FDD is pursuing a pro-activemore » solution to close the basins in-place and prevent a release to the groundwater. In-situ ion exchange is currently underway at the R-Reactor Disassembly Basin to reduce the Cs and Sr concentrations to levels that would allow release of the treated water to previously used on-site cooling ponds or to prevent ground water impact. The closure will be accomplished under CERCLA.« less

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.

METALLURGY DIVISION QUARTERLY REPORT FOR JULY, AUGUST, AND SEPTEMBER 1957

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

None

1958-10-01

Advanced Water Reactor Program. Three firings were made of initial closed-porosity fuel pellet bodies. Each firing coatained pellets of the composition 90 wt.% ThO/sub 2/-10 wt.%fl U0/sub 2/ with various additives and firing variables. Fast Power Breeder Reactor Program. To determine the potential usefulness of a Zr-5 wt. % Pu alloy, the fabricability of the alloy was tested. The manufacture of rod stock from which fuel and blanket elements for the Mark III loading of the EBR-1 were prcduced has been completed. The effect of irradiation on extruded and heat-treated U-2 wt.% Zr alloy for the EBR- 1 is reported.more » Fabrication procedures for making graphite-U/sub 3/O/sub 8/ test specimens for the TREAT Reactor were investigated. Advanced Engineering and Development. Ultrasonic bond tests were conducted on 590 EBR-1 Mark III blanket fuel elemeats. The blanket rods and part of the fuel rcds for the EBR-1 Mark III loading are being checked for cladding thickness by an eddy current system. Investigations of corrosionresistant Zr-Nb alloy were coatinued. Corrosion of MR alloys is being studied Ln support of the Mighty Mouse reactor program. Dynamic corrosion tests were performed on aluminum alloys, and results are included. Prcduction, Treatment, and Properties of Materials. The progress of the program of preparing highpurity Pu by fused salt electrolysis is summarized. Velocities of ultrasonic waves propagated in directions suitable for determining the room- temperature elastic moduli C/sub 12/, C/sub 13/, and C/sub 23/ of alpha U were determined. investigation of recrystallization in heavily coldrolled alpha- uranium sheet without a texture change was essentially concluded during this quarter. Selfdiffasion runs in polycrystalline uranium in the gamma phase, using the sputtering technique, have yielded a tentative value for the diffusion coefficient between 10/sup -8/ and 10/sup -7/ cm/sup 2/second. The preparation of high-purity U-Pan alloys is reponted. The data for the alpha-tobeta transformation temperatures in high-purity U and U-C alloys were confirmed by experiments on new specimens. Microstructure, density, and thermal arrest data were obtained for several injection cast, nominal U-5 wt.%fl fissium and U-8 wt.%fl fissium alloys. Phase diagrams are preseated for U-Mo and U-Ru alloys. Alloy Theory and The Nature of Solids. Four new isomorphs of Ti/sub 2/Ni have been discovered. Effects of Irradiation on Materials. The experimental and analytical work on the radial distribution of thermal neutrons within cylindrically shaped fuel specimens during irradiation was completed. (For preceding period see ANL-5790.) (W.L.H.)« less

Update on Small Modular Reactors Dynamic System Modeling Tool: Web Application

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

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

Previous reports focused on the development of component and system models as well as end-to-end system models using Modelica and Dymola for two advanced reactor architectures: (1) Advanced Liquid Metal Reactor and (2) fluoride high-temperature reactor (FHR). The focus of this report is the release of the first beta version of the web-based application for model use and collaboration, as well as an update on the FHR model. The web-based application allows novice users to configure end-to-end system models from preconfigured choices to investigate the instrumentation and controls implications of these designs and allows for the collaborative development of individualmore » component models that can be benchmarked against test systems for potential inclusion in the model library. A description of this application is provided along with examples of its use and a listing and discussion of all the models that currently exist in the library.« less

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2010-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2011-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2012-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

Development of a polysilicon process based on chemical vapor deposition, phase 1 and phase 2

NASA Technical Reports Server (NTRS)

Plahutnik, F.; Arvidson, A.; Sawyer, D.; Sharp, K.

1982-01-01

High-purity polycrystalline silicon was produced in an experimental, intermediate and advanced CVD reactor. Data from the intermediate and advanced reactors confirmed earlier results obtained in the experimental reactor. Solar cells were fabricated by Westinghouse Electric and Applied Solar Research Corporation which met or exceeded baseline cell efficiencies. Feedstocks containing trichlorosilane or silicon tetrachloride are not viable as etch promoters to reduce silicon deposition on bell jars. Neither are they capable of meeting program goals for the 1000 MT/yr plant. Post-run CH1 etch was found to be a reasonably effective method of reducing silicon deposition on bell jars. Using dichlorosilane as feedstock met the low-cost solar array deposition goal (2.0 gh-1-cm-1), however, conversion efficiency was approximately 10% lower than the targeted value of 40 mole percent (32 to 36% achieved), and power consumption was approximately 20 kWh/kg over target at the reactor.

Neutronics Analysis of SMART Small Modular Reactor using SRAC 2006 Code

NASA Astrophysics Data System (ADS)

Ramdhani, Rahmi N.; Prastyo, Puguh A.; Waris, Abdul; Widayani; Kurniadi, Rizal

2017-07-01

Small modular reactors (SMRs) are part of a new generation of nuclear reactor being developed worldwide. One of the advantages of SMR is the flexibility to adopt the advanced design concepts and technology. SMART (System integrated Modular Advanced ReacTor) is a small sized integral type PWR with a thermal power of 330 MW that has been developed by KAERI (Korea Atomic Energy Research Institute). SMART core consists of 57 fuel assemblies which are based on the well proven 17×17 array that has been used in Korean commercial PWRs. SMART is soluble boron free, and the high initial reactivity is mainly controlled by burnable absorbers. The goal of this study is to perform neutronics evaluation of SMART core with UO2 as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2006 code with JENDL 3.3 as nuclear data library.

Applying flow chemistry: methods, materials, and multistep synthesis.

PubMed

McQuade, D Tyler; Seeberger, Peter H

2013-07-05

The synthesis of complex molecules requires control over both chemical reactivity and reaction conditions. While reactivity drives the majority of chemical discovery, advances in reaction condition control have accelerated method development/discovery. Recent tools include automated synthesizers and flow reactors. In this Synopsis, we describe how flow reactors have enabled chemical advances in our groups in the areas of single-stage reactions, materials synthesis, and multistep reactions. In each section, we detail the lessons learned and propose future directions.

ADVANCED COURSE ON FUEL ELEMENTS FOR WATER COOLED POWER REACTORS, ORGANIZED BY THE NETHERLANDS'-NORWEGIAN REACTOR SCHOOL AT INSTITUTT FOR ATOMENERGI, KJELLER, NORWAY, 22nd AUGUST-3rd SEPTEMBER,1960. VOLUME III

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

Aas, S.; Barendregt, T.J.; Chesne, A.

1960-07-01

A series of lectures on fuel elements for water-cooled power reactors are presented. Topics covered include fabrication, properties, cladding, radiation damage, design, cycling, storage and transpont, and reprocessing. Separate records have been prepared for each section.

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.

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

Introduction to the special issue: 50th anniversary of APA Division 28: The past, present, and future of psychopharmacology and substance abuse.

PubMed

Stoops, William W; Sigmon, Stacey C; Evans, Suzette M

2016-08-01

This is an introduction to the special issue "50th Anniversary of APA Division 28: The Past, Present, and Future of Psychopharmacology and Substance Abuse." Taken together, the scholarly contributions included in this special issue serve as a testament to the important work conducted by our colleagues over the past five decades. Division 28 and its members have advanced and disseminated knowledge on the behavioral effects of drugs, informed efforts to prevent and treat substance abuse, and influenced education and policy issues more generally. As past and current leaders of the division, we are excited to celebrate 50 years of Division 28 and look forward to many more successful decades for our division and its members. (PsycINFO Database Record (c) 2016 APA, all rights reserved).

School of Advanced Military Studies Research Catalog AY 1983 - 1984 through 1991 - 1992

DTIC Science & Technology

1992-01-01

Battalion Night Attack- Command and Con- trol System, Strengths and Weaknesses, ADA Pierce, Kerry K ., Major, EN, E-Force: How 179 280 Agile Is It?, ADA 179...Sturgeon, Douglas E., Major, MI, Gaining an Pierce, Kerry K ., Major, EN, Kursk: A Study Operational Advantage: The Interdiction of in Operational Art...Synchronization Ability of Division? try Division METL and are the Divisions ADA 225 464 Training to Accomplish Them? ADA 225 483 Major James K . Greer, The

On Point: The United States Army in Operation Iraqi Freedom

DTIC Science & Technology

2004-01-01

arrived and prepared. The I MEF, commanding 1st Marine Division (1 MARDIV), 1 Air Wing, the 1st Armoured Division (UK), and other supporting units, made...reach into Baghdad. Supporting and Parallel Operations As V Corps advanced north toward Baghdad, I MEF, supported by the 1st (UK) Armoured Division...The 3rd ACR had weeks before it expected to enter the theater, as did the 2nd ACR (L). Three of the 7th UK Armoured Brigade’s four battle groups

23 CFR 635.115 - Agreement estimate.

Code of Federal Regulations, 2014 CFR

2014-04-01

... prepared by the STD and submitted to the Division Administrator as soon as practicable for use in the... submitted by the STD for each force account project (see 23 CFR part 635, subpart B) when the plans and... estimated quantities and the unit prices agreed upon in advance between the STD and the Division...

23 CFR 635.115 - Agreement estimate.

Code of Federal Regulations, 2011 CFR

2011-04-01

... prepared by the STD and submitted to the Division Administrator as soon as practicable for use in the... submitted by the STD for each force account project (see 23 CFR part 635, subpart B) when the plans and... estimated quantities and the unit prices agreed upon in advance between the STD and the Division...

23 CFR 635.115 - Agreement estimate.

Code of Federal Regulations, 2013 CFR

2013-04-01

... prepared by the STD and submitted to the Division Administrator as soon as practicable for use in the... submitted by the STD for each force account project (see 23 CFR part 635, subpart B) when the plans and... estimated quantities and the unit prices agreed upon in advance between the STD and the Division...

15 CFR 745.1 - Advance notification and annual report of all exports of Schedule 1 chemicals to other States...

Code of Federal Regulations, 2010 CFR

2010-01-01

... address: Information Technology Team, Treaty Compliance Division, Bureau of Industry and Security, U.S... courier delivery to the following address: Information Technology Team, Treaty Compliance Division, Bureau... Foreign Trade Regulations Relating to Commerce and Foreign Trade (Continued) BUREAU OF INDUSTRY AND...

Effective Teaching in Elementary Social Studies.

ERIC Educational Resources Information Center

Savage, Tom V.; Armstrong, David G.

This book is designed for use in elementary social studies methods classes, as a source for discussion in advanced curriculum classes, and as a personal reference for elementary social studies teachers. This book has four major divisions with each division offering a list of lesson ideas. Part 1, "Contexts for the Social Studies," includes: (1)…

Cermet-fueled reactors for advanced space applications

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

Cowan, C.L.; Palmer, R.S.; Taylor, I.N.

Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel weremore » carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper.« less

A summary of sodium-cooled fast reactor development

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

Aoto, Kazumi; Dufour, Philippe; Hongyi, Yang

Much of the basic technology for the Sodium-cooled fast Reactor (SFR) has been established through long term development experience with former fast reactor programs, and is being confirmed by the Phénix end-of-life tests in France, the restart of Monju in Japan, the lifetime extension of BN-600 in Russia, and the startup of the China Experimental Fast Reactor in China. Planned startup in 2014 for new SFRs: BN-800 in Russia and PFBR in India, will further enhance the confirmation of the SFR basic technology. Nowadays, the SFR development has advanced to aiming at establishment of the Generation-IV system which is dedicatedmore » to sustainable energy generation and actinide management, and several advanced SFR concepts are under development such as PRISM, JSFR, ASTRID, PGSFR, BN-1200, and CFR-600. Generation-IV International Forum is an international collaboration framework where various R&D activities are progressing on design of system and component, safety and operation, advanced fuel, and actinide cycle for the Generation-IV SFR development, and will play a beneficial role of promoting them thorough providing an opportunity to share the past experience and the latest data of design and R&D among countries developing SFR.« less

PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2

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

Hunn, John D.; Gerczak, Tyler J.; Morris, Robert Noel

2016-04-01

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in themore » Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.« less

Technology transfer in the NASA Ames Advanced Life Support Division

NASA Technical Reports Server (NTRS)

Connell, Kathleen; Schlater, Nelson; Bilardo, Vincent; Masson, Paul

1992-01-01

This paper summarizes a representative set of technology transfer activities which are currently underway in the Advanced Life Support Division of the Ames Research Center. Five specific NASA-funded research or technology development projects are synopsized that are resulting in transfer of technology in one or more of four main 'arenas:' (1) intra-NASA, (2) intra-Federal, (3) NASA - aerospace industry, and (4) aerospace industry - broader economy. Each project is summarized as a case history, specific issues are identified, and recommendations are formulated based on the lessons learned as a result of each project.

Communications and Intelligent Systems Division Overview

NASA Technical Reports Server (NTRS)

Emerson, Dawn

2016-01-01

This presentation provides an overview of the research and engineering in the competency fieldsof advanced communications and intelligent systems with emphasis on advanced technologies, architecture definitionand system development for application in current and future aeronautics and space systems.

Generalized Nanosatellite Avionics Testbed Lab

NASA Technical Reports Server (NTRS)

Frost, Chad R.; Sorgenfrei, Matthew C.; Nehrenz, Matt

2015-01-01

The Generalized Nanosatellite Avionics Testbed (G-NAT) lab at NASA Ames Research Center provides a flexible, easily accessible platform for developing hardware and software for advanced small spacecraft. A collaboration between the Mission Design Division and the Intelligent Systems Division, the objective of the lab is to provide testing data and general test protocols for advanced sensors, actuators, and processors for CubeSat-class spacecraft. By developing test schemes for advanced components outside of the standard mission lifecycle, the lab is able to help reduce the risk carried by advanced nanosatellite or CubeSat missions. Such missions are often allocated very little time for testing, and too often the test facilities must be custom-built for the needs of the mission at hand. The G-NAT lab helps to eliminate these problems by providing an existing suite of testbeds that combines easily accessible, commercial-offthe- shelf (COTS) processors with a collection of existing sensors and actuators.

Transuranic inventory reduction in repository by partitioning and transmutation

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

Kang, C.H.; Kazimi, M.S.

1992-01-01

The promise of a new reprocessing technology and the issuance of Environmental Protection Agency (EPA) and U.S. Nuclear Regulatory Commission regulations concerning a geologic repository rekindle the interest in partitioning and transmutation of transuranic (TRU) elements from discharged reactor fuel as a high level waste management option. This paper investigates the TRU repository inventory reduction capability of the proposed advanced liquid metal reactors (ALMRs) and integral fast reactors (IFRs) as well as the plutonium recycled light water reactors (LWRs).

Fuel Fabrication and Nuclear Reactors

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

Karpius, Peter Joseph

2017-02-02

The uranium from the enrichment plant is still in the form of UF 6. UF 6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF 6 is converted into UO 2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

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

A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

2010-06-01

2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has beenmore » restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.« less

Princeton Plasma Physics Laboratory: Annual report, October 1, 1986--September 30, 1987

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

Not Available

1987-01-01

This report contains papers on the following topics: Principle Parameters Achieved in Experimental Devices (FY87); Tokamak Fusion Test Reactor; Princeton Beta Experiment-Modification; S-1 Spheromak; Current-Drive Experiment; X-Ray Laser Studies; Theoretical Division; Tokamak Modeling; Compact Ignition Tokamak; Engineering Department; Project Planning and Safety Office; Quality Assurance and Reliability; Administrative Operations; and PPPL Patent Invention Disclosures (FY87).

16. ARAII Administration building ARA613. South (front) and east sides. ...

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

16. ARA-II Administration building ARA-613. South (front) and east sides. Camera facing northwest. Sign at left corner of building says, "Fuels and materials division, materials joining research and development laboratory." Part of south wall already has been demolished. Sign on roof railing says, "Danger--Abestos." Ineel photo no. 2-3. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Neutron Physics Division progress report for period ending February 28, 1977

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

Maienschein, F.C.

1977-05-01

Summaries are given of research progress in the following areas: (1) measurements of cross sections and related quantities, (2) cross section evaluations and theory, (3) cross section processing, testing, and sensitivity analysis, (4) integral experiments and their analyses, (5) development of methods for shield and reactor analyses, (6) analyses for specific systems or applications, and (7) information analysis and distribution. (SDF)

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

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

Smith, Kurt R.; Howard, Richard H.; Daily, Charles R.

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designsmore » allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.« less

D-He-3 spherical torus fusion reactor system study

NASA Astrophysics Data System (ADS)

Macon, William A., Jr.

1992-04-01

This system study extrapolates present physics knowledge and technology to predict the anticipated characteristics of D-He3 spherical torus fusion reactors and their sensitivity to uncertainties in important parameters. Reference cases for steady-state 1000 MWe reactors operating in H-mode in both the 1st stability regime and the 2nd stability regime were developed and assessed quantitatively. These devices would a very small aspect ratio (A=1,2), a major radius of about 2.0 m, an on-axis magnetic field less than 2 T, a large plasma current (80-120 MA) dominated by the bootstrap effect, and high plasma beta (greater than O.6). The estimated cost of electricity is in the range of 60-90 mills/kW-hr, assuming the use of a direct energy conversion system. The inherent safety and environmental advantages of D-He3 fusion indicate that this reactor concept could be competitive with advanced fission breeder reactors and large-scale solar electric plants by the end of the 21st century if research and development can produce the anticipated physics and technology advances.

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

Gerczak, Tyler J.; Smith, Kurt R.; Petrie, Christian M.

Tristructural-isotropic (TRISO)–coated particle fuel is a promising advanced fuel concept consisting of a spherical fuel kernel made of uranium oxide and uranium carbide, surrounded by a porous carbonaceous buffer layer and successive layers of dense inner pyrolytic carbon (IPyC), silicon carbide (SiC) deposited by chemical vapor , and dense outer pyrolytic carbon (OPyC). This fuel concept is being considered for advanced reactor applications such as high temperature gas-cooled reactors (HTGRs) and molten salt reactors (MSRs), as well as for accident-tolerant fuel for light water reactors (LWRs). Development and implementation of TRISO fuel for these reactor concepts support the US Departmentmore » of Energy (DOE) Office of Nuclear Energy mission to promote safe, reliable nuclear energy that is sustainable and environmentally friendly. During operation, the SiC layer serves as the primary barrier to metallic fission products and actinides not retained in the kernel. It has been observed that certain fission products are released from TRISO fuel during operation, notably, Ag, Eu, and Sr [1]. Release of these radioisotopes causes safety and maintenance concerns.« less

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

Incorporation of real-time component information using equipment condition assessment (ECA) through the developmentof enhanced risk monitors (ERM) for active components in advanced reactor (AR) and advanced small modular reactor (SMR) designs. We incorporate time-dependent failure probabilities from prognostic health management (PHM) systems to dynamically update the risk metric of interest. This information is used to augment data used for supervisory control and plant-wide coordination of multiple modules by providing the incremental risk incurred due to aging and demands placed on components that support mission requirements.

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

Natesan, K.; Momozaki, Y.; Li, M.

This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory,more » the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux Test Facility, and Clinch River Breeder Reactor. Among the nonmetallic elements discussed, oxygen is deemed controllable and its concentration in sodium can be maintained in sodium for long reactor life by using cold-trap method. It was concluded that among the cold-trap and getter-trap methods, the use of cold trap is sufficient to achieve oxygen concentration of the order of 1 part per million. Under these oxygen conditions in sodium, the corrosion performance of structural materials such as austenitic stainless steels and ferritic steels will be acceptable at a maximum core outlet sodium temperature of {approx}550 C. In the current sodium compatibility studies, the oxygen concentration in sodium will be controlled and maintained at {approx}1 ppm by controlling the cold trap temperature. The oxygen concentration in sodium in the forced convection sodium loop will be controlled and monitored by maintaining the cold trap temperature in the range of 120-150 C, which would result in oxygen concentration in the range of 1-2 ppm. Uniaxial tensile specimens are being exposed to flowing sodium and will be retrieved and analyzed for corrosion and post-exposure tensile properties. Advanced materials for sodium exposure include austenitic alloy HT-UPS and ferritic-martensitic steels modified 9Cr-1Mo and NF616. Among the nonmetallic elements in sodium, carbon was assessed to have the most influence on structural materials since carbon, as an impurity, is not amenable to control and maintenance by any of the simple purification methods. The dynamic equilibrium value for carbon in sodium systems is dependent on several factors, details of which were discussed in the earlier report. The current sodium compatibility studies will examine the role of carbon concentration in sodium on the carburization-decarburization of advanced structural materials at temperatures up to 650 C. Carbon will be added to the sodium by exposure of carbon-filled iron tubes, which over time will enable carbon to diffuse through iron and dissolve into sodium. The method enables addition of dissolved carbon (without carbon particulates) in sodium that is of interest for materials compatibility evaluation. The removal of carbon from the sodium will be accomplished by exposing carbon-gettering alloys such as refractory metals that have a high partitioning coefficient for carbon and also precipitate carbides, thereby decreasing the carbon concentration in sodium.« less

Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

DTIC Science & Technology

2006-12-01

24 Table 3.3 Hazards of Sodium Reaction Products, Hydride And Oxide...........................26 Table 3.4 Chemical Reactivity Of Selected...Liquid Metal Fast Breeder Reactor ORIGEN Oak Ridge Isotope Generator ORIGENARP Oak Ridge Isotope Generator Automated Rapid Processing PWR ...nuclear reactors, both because of the possibility of increased reactivity due to boiling and the potential loss of effectiveness of coolant heat transfer

Syngas Production By Thermochemical Conversion Of H2o And Co2 Mixtures Using A Novel Reactor Design

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

Pearlman, Howard; Chen, Chien-Hua

The Department of Energy awarded Advanced Cooling Technologies, Inc. (ACT) an SBIR Phase II contract (#DE-SC0004729) to develop a high-temperature solar thermochemical reactor for syngas production using water and/or carbon dioxide as feedstocks. The technology aims to provide a renewable and sustainable alternative to fossil fuels, promote energy independence and mitigate adverse issues associated with climate change by essentially recycling carbon from carbon dioxide emitted by the combustion of hydrocarbon fuels. To commercialize the technology and drive down the cost of solar fuels, new advances are needed in materials development and reactor design, both of which are integral elements inmore » this program.« 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

Academic and Career Advancement for Black Male Athletes at NCAA Division I Institutions

ERIC Educational Resources Information Center

Baker, Ashley R.; Hawkins, Billy J.

2016-01-01

This chapter examines the structural arrangements and challenges many Black male athletes encounter as a result of their use of sport for upward social mobility. Recommendations to enhance their preparation and advancement are provided.

University Teaching and Social Cohesion in the Age of AIDS: A South African Case Study

ERIC Educational Resources Information Center

Lesko, Nancy

2007-01-01

In the context of ongoing social divisions, lack of coherent leadership by government, and even divisiveness over medical advances and public health mandates, how might universities respond? What university actions can support "social cohesion" in a society splintered by class, race, gender, colonial legacies, the history of apartheid,…

Early Childhood Special Education and Early Intervention Personnel Preparation Standards of the Division for Early Childhood: Field Validation

ERIC Educational Resources Information Center

Cochran, Deborah C.; Gallagher, Peggy A.; Stayton, Vicki D.; Dinnebeil, Laurie A.; Lifter, Karin; Chandler, Lynette K.; Christensen, Kimberly A.

2012-01-01

Results of the field validation survey of the revised initial and new advanced Council for Exceptional Children (CEC) Division for Early Childhood (DEC) early childhood special education (ECSE)/early intervention (EI) personnel standards are presented. Personnel standards are used as part of educational accountability systems and in teacher…

Advanced Small Modular Reactor Economics Status Report

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

Harrison, Thomas J.

2014-10-01

This report describes the data collection work performed for an advanced small modular reactor (AdvSMR) economics analysis activity at the Oak Ridge National Laboratory. The methodology development and analytical results are described in separate, stand-alone documents as listed in the references. The economics analysis effort for the AdvSMR program combines the technical and fuel cycle aspects of advanced (non-light water reactor [LWR]) reactors with the market and production aspects of SMRs. This requires the collection, analysis, and synthesis of multiple unrelated and potentially high-uncertainty data sets from a wide range of data sources. Further, the nature of both economic andmore » nuclear technology analysis requires at least a minor attempt at prediction and prognostication, and the far-term horizon for deployment of advanced nuclear systems introduces more uncertainty. Energy market uncertainty, especially the electricity market, is the result of the integration of commodity prices, demand fluctuation, and generation competition, as easily seen in deregulated markets. Depending on current or projected values for any of these factors, the economic attractiveness of any power plant construction project can change yearly or quarterly. For long-lead construction projects such as nuclear power plants, this uncertainty generates an implied and inherent risk for potential nuclear power plant owners and operators. The uncertainty in nuclear reactor and fuel cycle costs is in some respects better understood and quantified than the energy market uncertainty. The LWR-based fuel cycle has a long commercial history to use as its basis for cost estimation, and the current activities in LWR construction provide a reliable baseline for estimates for similar efforts. However, for advanced systems, the estimates and their associated uncertainties are based on forward-looking assumptions for performance after the system has been built and has achieved commercial operation. Advanced fuel materials and fabrication costs have large uncertainties based on complexities of operation, such as contact-handled fuel fabrication versus remote handling, or commodity availability. Thus, this analytical work makes a good faith effort to quantify uncertainties and provide qualifiers, caveats, and explanations for the sources of these uncertainties. The overall result is that this work assembles the necessary information and establishes the foundation for future analyses using more precise data as nuclear technology advances.« less

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

Harmony, S.C.; Steiner, J.L.; Stumpf, H.J.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. As part of the preapplication and eventual design certification process, advanced reactor applicants are required to submit neutronic and thermal-hydraulic safety analyses over a sufficient range of normal operation, transient conditions, and specified accident sequences. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. A fully one-dimensional modelmore » of the PIUS reactor has been developed for the Transient Reactor Analysis Code, TRACPF1/MOD2. Early in 1992, ABB submitted a Supplemental Information Package describing recent design modifications. An important feature of the PIUS Supplement design was the addition of an active scram system that will function for most transient and accident conditions. A one-dimensional Transient Reactor Analysis Code baseline calculation of the PIUS Supplement design were performed for a break in the main steam line at the outlet nozzle of the loop 3 steam generator. Sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions following a main steam line break. The sensitivity study results provide insights into the robustness of the design.« less

Advanced Electron Microscopy and Micro analytical technique development and application for Irradiated TRISO Coated Particles from the AGR-1 Experiment

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

Van Rooyen, Isabella Johanna; Lillo, Thomas Martin; Wen, Haiming

2017-01-01

A series of up to seven irradiation experiments are planned for the Advanced Gas Reactor (AGR) Fuel Development and Quantification Program, with irradiation completed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for the first experiment (i.e., AGR-1) in November 2009 for an effective 620 full power days. The objective of the AGR-1 experiment was primarily to provide lessons learned on the multi-capsule test train design and to provide early data on fuel performance for use in fuel fabrication process development and post-irradiation safety testing data at high temperatures. This report describes the advanced microscopy and micro-analysismore » results on selected AGR-1 coated particles.« less

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

10 CFR 110.41 - Executive Branch review.

Code of Federal Regulations, 2011 CFR

2011-01-01

.... (6) An export involving assistance to end uses related to isotope separation, chemical reprocessing, heavy water production, advanced reactors, or the fabrication of nuclear fuel containing plutonium... equipment to a foreign reactor. (8) An export involving radioactive waste. (9) An export to any country...

Cognitive and Neural Sciences Division 1990 Programs

DTIC Science & Technology

1990-08-01

operator skill acquisition for advanced manufacturing environments. In (W. Karawowski and M. Rahimi, Eds.) Ergonomics of Advanced Manufacturing and...PRINCIPAL INVESTIGATOR: John H. Maunsell University of Rochester Strong School of Medicine and Dentistry (716) 275-2076 R&T PROJECT CODE: 4424242 CONTRACT

Magnet design with 100-kA HTS STARS conductors for the helical fusion reactor

NASA Astrophysics Data System (ADS)

Yanagi, N.; Terazaki, Y.; Ito, S.; Tamura, H.; Hamaguchi, S.; Mito, T.; Hashizume, H.; Sagara, A.

2016-12-01

The high-temperature superconducting (HTS) option is employed for the conceptual design of the LHD-type helical fusion reactor FFHR-d1. The 100-kA-class STARS (Stacked Tapes Assembled in Rigid Structure) conductor is used for the magnet system including the continuously wound helical coils. Protection of the magnet system in case of a quench is a crucial issue and the hot-spot temperature during an emergency discharge is estimated based on the zero-dimensional and one-dimensional analyses. The number of division of the coil winding package is examined to limit the voltage generation. For cooling the HTS magnet, helium gas flow is considered and its feasibility is examined by simple analysis as a first step.

DEMINERALIZER BUILDING, TRA608. CAMERA IS ON RAW WATER TOWER AND ...

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

DEMINERALIZER BUILDING, TRA-608. CAMERA IS ON RAW WATER TOWER AND FACES WEST. STEAM PLANT, TRA-609, AT UPPER EDGE OF VIEW. ABSENCE OF ROOF EXPOSES FIVE-BAY STRUCTURE AND INTERIOR DIVISION OF SPACE. CORRIDOR AT WEST END OF BUILDING WILL SEPARATE LABORATORY AND OFFICE SPACE FROM POTABLE WATER TANKS. ALONG NORTH WALL ARE SPACES FOR CATION AND ANION EXCHANGE UNITS. PENTHOUSE WILL ENCLOSE DEGASSIFIER. TANK AT LEFT (SOUTH) OF BUILDING STORES DEMINERALIZED WATER. NOTE BRINE STORAGE PIT, TRA-631, AT RIGHT OF VIEW, ABOVE PAIR OF CAUSTIC STORAGE TANKS. NOTE TRENCHES FOR BURIED WATER PIPES. INL NEGATIVE NO. 2732. Unknown Photographer, 6/29/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Advanced Reactors-Intermediate Heat Exchanger (IHX) Coupling: Theoretical Modeling and Experimental Validation

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

Utgikar, Vivek; Sun, Xiaodong; Christensen, Richard

2016-12-29

The overall goal of the research project was to model the behavior of the advanced reactorintermediate heat exchange system and to develop advanced control techniques for off-normal conditions. The specific objectives defined for the project were: 1. To develop the steady-state thermal hydraulic design of the intermediate heat exchanger (IHX); 2. To develop mathematical models to describe the advanced nuclear reactor-IHX-chemical process/power generation coupling during normal and off-normal operations, and to simulate models using multiphysics software; 3. To develop control strategies using genetic algorithm or neural network techniques and couple these techniques with the multiphysics software; 4. To validate themore » models experimentally The project objectives were accomplished by defining and executing four different tasks corresponding to these specific objectives. The first task involved selection of IHX candidates and developing steady state designs for those. The second task involved modeling of the transient and offnormal operation of the reactor-IHX system. The subsequent task dealt with the development of control strategies and involved algorithm development and simulation. The last task involved experimental validation of the thermal hydraulic performances of the two prototype heat exchangers designed and fabricated for the project at steady state and transient conditions to simulate the coupling of the reactor- IHX-process plant system. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed high-temperature molten salt facility.« less

Summary of space nuclear reactor power systems, 1983--1992

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

Buden, D.

1993-08-11

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressedmore » from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.« less

FFTF Passive Safety Test Data for Benchmarks for New LMR Designs

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

Wootan, David W.; Casella, Andrew M.

Liquid Metal Reactors (LMRs) continue to be considered as an attractive concept for advanced reactor design. Software packages such as SASSYS are being used to im-prove new LMR designs and operating characteristics. Significant cost and safety im-provements can be realized in advanced liquid metal reactor designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associ-ated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. The FFTF passive safety testing pro-gram was developed to examine howmore » specific design elements influenced dynamic re-activity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results to reactors of current interest. The U.S. Department of En-ergy, Office of Nuclear Energy Advanced Reactor Technology program is in the pro-cess of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Benchmarks based on empirical data gathered during operation of the Fast Flux Test Facility (FFTF) as well as design documents and post-irradiation examination will aid in the validation of these software packages and the models and calculations they produce. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs« less

Summary of space nuclear reactor power systems, 1983 - 1992

NASA Astrophysics Data System (ADS)

Buden, D.

1993-08-01

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987-88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

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

Hu, Rui

2017-09-03

Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developedmore » at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.« less

The Nuclear Renaissance — Implications on Quantitative Nondestructive Evaluations

NASA Astrophysics Data System (ADS)

Matzie, Regis A.

2007-03-01

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

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

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

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

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

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

DOE PAGES

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

2017-02-26

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

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

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

S. T. Khericha

2007-04-01

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

Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

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

Bond, Leonard J.; Bowler, John R.

The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-servicemore » inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO 3-xPbTiO 3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.« less

NASA Exploration Forum: Human Path to Mars

NASA Image and Video Library

2014-04-29

Sam Scimemi, Director of NASA's International Space Station Division, second from left, Phil McAlister, Director of NASA's Commercial Spaceflight Division, third from left, Dan Dumbacher, Deputy Associate Administrator of NASA's Exploration Systems Development, center, Michele Gates, Senior Technical Advisor of NASA's Human Exploration and Operations Mission Directorate, second from right, and Jason Crusan, Director of NASA's Advanced Exploration Systems Division, right, sit on a panel during an Exploration Forum showcasing NASA's human exploration path to Mars in the James E. Webb Auditorium at NASA Headquarters on Tuesday, April 29, 2014. Photo Credit: (NASA/Joel Kowsky)

NASA Exploration Forum: Human Path to Mars

NASA Image and Video Library

2014-04-29

Sam Scimemi, Director of NASA's International Space Station Division, left, Phil McAlister, Director of NASA's Commercial Spaceflight Division, second from left, Dan Dumbacher, Deputy Associate Administrator of NASA's Exploration Systems Development, center, Michele Gates, Senior Technical Advisor of NASA's Human Exploration and Operations Mission Directorate, second from right, and Jason Crusan, Director of NASA's Advanced Exploration Systems Division, right, sit on a panel during an Exploration Forum showcasing NASA's human exploration path to Mars in the James E. Webb Auditorium at NASA Headquarters on Tuesday, April 29, 2014. Photo Credit: (NASA/Joel Kowsky)

Cell cycles and cell division in the archaea.

PubMed

Samson, Rachel Y; Bell, Stephen D

2011-06-01

Until recently little was known about the cell cycle parameters and division mechanisms of archaeal organisms. Although this is still the case for the majority of archaea, significant advances have been made in some model species. The information that has been gleaned thus far points to a remarkable degree of diversity within the archaeal domain of life. More specifically, members of distinct phyla have very different chromosome copy numbers, replication control systems and even employ distinct machineries for cell division. Copyright © 2011 Elsevier Ltd. All rights reserved.

Risk Management for Sodium Fast Reactors.

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

Denman, Matthew R.; Groth, Katrina; Cardoni, Jeffrey N.

2015-01-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event withmore » the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.« less

Research and technology, 1990

NASA Technical Reports Server (NTRS)

1990-01-01

Selected research and technology activities at Ames Research Center, including the Moffett Field site and the Dryden Flight Research Facility, are summarized. These accomplishments exemplify the Center's varied and highly productive research efforts for 1990. The activities addressed are under the directories of: (1) aerospace systems which contains aircraft technology, full-scale aerodynamics research, information sciences, aerospace human factors research, and flight systems and simulation research divisions; (2) Dryden flight research facility which contains research engineering division; (3) aerophysics which contains aerodynamics, fluid dynamics, and thermosciences divisions; and (4) space research which contains advanced life support, space projects, earth system science, life science, and space science divisions, and search for extraterrestrial intelligence and space life sciences payloads offices.

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

Plant maintenance and advanced reactors issue, 2008

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

Agnihotri, Newal

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada;more » Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.« less

Preliminary design and hazards report. Boiling Reactor Experiment V (BORAX V)

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

Rice, R. E.

1960-02-01

The preliminary objectives of the proposed BORAX V program are to test nuclear superheating concepts and to advance the technology of boiling-water-reactor design by performing experiments which will improve the understanding of factors limiting the stability of boiling reactors at high power densities. The reactor vessel is a cylinder with ellipsoidal heads, made of carbon steel clad internally with stainless steel. Each of the three cores is 24 in. high and has an effective diameter of 39 in. This is a preliminary report. (W.D.M.)

KSC-08pd0982

NASA Image and Video Library

2008-04-18

CAPE CANAVERAL, Fla. -- During a break in presentations at NASA’s Future Forum in Miami, astronaut Steve Frick and former astronaut Carl Walz sign autographs. Frick served as commander of the STS-122 shuttle mission. Walz is now director of NASA’s Advanced Capabilities Division. The forum focused on how space exploration benefits Florida's economy. The event, which included presentations and panels, was held at the University of Miami's BankUnited Center. Among those participating were NASA Deputy Administrator Shana Dale, astronaut Carl Walz, director of the Advanced Capabilities Division in NASA's Exploration Systems Mission Directorate, and Russell Romanella, director, International Space Station and Spacecraft Processing. Photo credit: NASA/Kim Shiflett

Nano Goes Magnetic to Attract Big Business

NASA Technical Reports Server (NTRS)

2006-01-01

Glenn Research Center has combined state-of-the-art electrical designs with complex, computer-aided analyses to develop some of today s most advanced power systems, in space and on Earth. The center s Power and On-Board Propulsion Technology Division is the brain behind many of these power systems. For space, this division builds technologies that help power the International Space Station, the Hubble Space Telescope, and Earth-orbiting satellites. For Earth, it has woven advanced aerospace power concepts into commercial energy applications that include solar and nuclear power generation, battery and fuel cell energy storage, communications and telecommunications satellites, cryocoolers, hybrid and electric vehicles, and heating and air-conditioning systems.

ALARA Council: Sharing our resources and experiences to reduce doses at Commonwealth Edison Facilities

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

Rescek, F.

1995-03-01

Commonwealth Edison Company is an investor-owned utility company supplying electricity to over three million customers (eight million people) in Chicago and northern Illinois, USA. The company operates 16 generating stations which have the capacity to produce 22,522 megawatts of electricity. Six of these generating stations, containing 12 nuclear units, supply 51% of this capacity. The 12 nuclear units are comprised of four General Electric boiling water (BWR-3) reactors, two General Electric BWR-5 reactors, and six Westinghouse four-loop pressurized water reactors (PWR). In August 1993, Commonwealth Edison created an ALARA Council with the responsibility to provide leadership and guidance that resultsmore » in an effective ALARA Culture within the Nuclear Operations Division. Unlike its predecessor, the Corporate ALARA Committee, the ALARA Council is designed to bring together senior managers from the six nuclear stations and corporate to create a collaborative effort to reduce occupational doses at Commonwealth Edison`s stations.« less

U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development

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

George W. Griffith

2011-10-01

A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear Fuels Pathway. The intent of this work is to invest in a high-risk, high-reward technology that can be introduced in a relatively short time. The LWRS goal is to demonstrate successful advanced fuels technology that suitable for commercial development to support nuclear relicensing. Ceramic matrix composites are an established non-nuclear technology that utilizes ceramic fibers embedded in a ceramic matrix. A thin interfacial layer between the fibers and the matrix allows formore » ductile behavior. The SiC CMC has relatively high strength at high reactor accident temperatures when compared to metallic cladding. SiC also has a very low chemical reactivity and doesn't react exothermically with the reactor cooling water. The radiation behavior of SiC has also been studied extensively as structural fusion system components. The SiC CMC technology is in the early stages of development and will need to mature before confidence in the developed designs can created. The advanced SiC CMC materials do offer the potential for greatly improved safety because of their high temperature strength, chemical stability and reduced hydrogen generation.« less

Quarterly Progress Report for the Chemical and Energy Research Section of the Chemical Technology Division: July-September 1999

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

Jubin, R.T.

2001-04-16

This report summarizes the major activities conducted in the Chemical and Energy Research Section of the Chemical Technology Division at Oak Ridge National Laboratory (ORNL) during the period July-September 1999. The section conducts basic and applied research and development in chemical engineering, applied chemistry, and bioprocessing, with an emphasis on energy-driven technologies and advanced chemical separations for nuclear and waste applications. The report describes the various tasks performed within ten major areas of research: Hot Cell Operations, Process Chemistry, Molten Salt Reactor Experiment (MSRE) Remediation Studies, Chemistry Research, Physical Properties Research, Biochemical Engineering, Separations and Materials Synthesis, Fluid Structures andmore » Properties, Biotechnology Research, and Molecular Studies. The name of a technical contact is included with each task described, and readers are encouraged to contact these individuals if they need additional information. Activities conducted within the area of the Cell Operations involved the testing of two continuously stirred tank reactors in series to evaluate the Savannah River-developed process of small-tank tetraphenylborate precipitation to remove cesium, strontium and transuranics from supernatant. Within the area of Process Chemistry, various topics related to solids formation in process solutions from caustic treatment of Hanford sludge were addressed. Saltcake dissolution efforts continued, including the development of a predictive algorithm. New initiatives for the section included modeling activities centered on detection of hydrogen in {sup 233}U storage wells and wax formation in petroleum mixtures, as well as support for the Spallation Neutron Source (investigation of transmutation products formed during operation). Other activities involved in situ grouting and evaluation of options for use (i.e., as castable shapes) of depleted uranium. In a continuation of activities of the preceding quarter, MSRE Remediation Studies focused on recovery of {sup 233}U and its conversion to a stable oxide and radiolysis experiments to permit remediation of MSRE fuel salt. Investigation of options for final disposition of the {sup 233}U inventory represents a new initiative within this area. In the area of Chemistry Research, activities included studies relative to molecular imprinting for use in areas such as selective sorption, chemical sensing, and catalysis, as well as spectroscopic investigation into the fundamental interaction between ionic solvents and solutes in both low- and high-temperature ionic liquids.« less

Advances in boron neutron capture therapy (BNCT) at kyoto university - From reactor-based BNCT to accelerator-based BNCT

NASA Astrophysics Data System (ADS)

Sakurai, Yoshinori; Tanaka, Hiroki; Takata, Takushi; Fujimoto, Nozomi; Suzuki, Minoru; Masunaga, Shinichiro; Kinashi, Yuko; Kondo, Natsuko; Narabayashi, Masaru; Nakagawa, Yosuke; Watanabe, Tsubasa; Ono, Koji; Maruhashi, Akira

2015-07-01

At the Kyoto University Research Reactor Institute (KURRI), a clinical study of boron neutron capture therapy (BNCT) using a neutron irradiation facility installed at the research nuclear reactor has been regularly performed since February 1990. As of November 2014, 510 clinical irradiations were carried out using the reactor-based system. The world's first accelerator-based neutron irradiation system for BNCT clinical irradiation was completed at this institute in early 2009, and the clinical trial using this system was started in 2012. A shift of BCNT from special particle therapy to a general one is now in progress. To promote and support this shift, improvements to the irradiation system, as well as its preparation, and improvements in the physical engineering and the medical physics processes, such as dosimetry systems and quality assurance programs, must be considered. The recent advances in BNCT at KURRI are reported here with a focus on physical engineering and medical physics topics.

Advances in algal-prokaryotic wastewater treatment: A review of nitrogen transformations, reactor configurations and molecular tools.

PubMed

Wang, Meng; Keeley, Ryan; Zalivina, Nadezhda; Halfhide, Trina; Scott, Kathleen; Zhang, Qiong; van der Steen, Peter; Ergas, Sarina J

2018-07-01

The synergistic activity of algae and prokaryotic microorganisms can be used to improve the efficiency of biological wastewater treatment, particularly with regards to nitrogen removal. For example, algae can provide oxygen through photosynthesis needed for aerobic degradation of organic carbon and nitrification and harvested algal-prokaryotic biomass can be used to produce high value chemicals or biogas. Algal-prokaryotic consortia have been used to treat wastewater in different types of reactors, including waste stabilization ponds, high rate algal ponds and closed photobioreactors. This review addresses the current literature and identifies research gaps related to the following topics: 1) the complex interactions between algae and prokaryotes in wastewater treatment; 2) advances in bioreactor technologies that can achieve high nitrogen removal efficiencies in small reactor volumes, such as algal-prokaryotic biofilm reactors and enhanced algal-prokaryotic treatment systems (EAPS); 3) molecular tools that have expanded our understanding of the activities of algal and prokaryotic communities in wastewater treatment processes. Copyright © 2018 Elsevier Ltd. All rights reserved.

Communications and Intelligent Systems Division Overview

NASA Technical Reports Server (NTRS)

Emerson, Dawn

2017-01-01

This presentation provides an overview of the research and engineering work being performed in the competency fields of advanced communications and intelligent systems with emphasis on advanced technologies, architecture definition, and systems development for application in current and future aeronautics and space communications systems.

Report on International Education Review

ERIC Educational Resources Information Center

Ministry of Advanced Education, 2005

2005-01-01

The Ministry of Advanced Education (Ministry) provides overall funding and policy direction for British Columbia's public post-secondary education system. The Student and Strategic Services Division, Ministry of Advanced Education requested Internal Audit & Advisory Services to conduct a review of the international education programs at…

Thermodynamic analysis of the advanced zero emission power plant

NASA Astrophysics Data System (ADS)

Kotowicz, Janusz; Job, Marcin

2016-03-01

The paper presents the structure and parameters of advanced zero emission power plant (AZEP). This concept is based on the replacement of the combustion chamber in a gas turbine by the membrane reactor. The reactor has three basic functions: (i) oxygen separation from the air through the membrane, (ii) combustion of the fuel, and (iii) heat transfer to heat the oxygen-depleted air. In the discussed unit hot depleted air is expanded in a turbine and further feeds a bottoming steam cycle (BSC) through the main heat recovery steam generator (HRSG). Flue gas leaving the membrane reactor feeds the second HRSG. The flue gas consist mainly of CO2 and water vapor, thus, CO2 separation involves only the flue gas drying. Results of the thermodynamic analysis of described power plant are presented.

Young Kim, PhD | Division of Cancer Prevention

Cancer.gov

Young S Kim, PhD, joined the Division of Cancer Prevention at the National Cancer Institute in 1998 as a Program Director who oversees and monitors NCI grants in the area of Nutrition and Cancer. She serves as an expert in nutrition, molecular biology, and genomics as they relate to cancer prevention. Dr. Kim assists with research initiatives that will advance nutritional

US Army And The Emergence Of Unmanned Threats

DTIC Science & Technology

2016-05-26

43 iv Acronyms A2/AD Anti-Access/Area Denial AAA Anti-aircraft Artillery ADP Army...possessed sufficient resources and motivation to challenge local air superiority at the bridgehead. 66 Allied anti-aircraft artillery ( AAA ) raced to...Remagen. 482nd AAA , organic to Ninth Armored Division advanced through the division column to emplace at the bridge. With the vulnerability of the

Summaries of FY 1979 research in the chemical sciences

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

Not Available

1980-05-01

The purpose of this report is to help those interested in research supported by the Department of Energy's Division of Chemical Sciences, which is one of six Divisions of the Office of Basic Energy Sciences in the Office of Energy Research. Chemists, physicists, chemical engineers and others who are considering the possibility of proposing research for support by this Division wll find the booklet useful for gauging the scope of the program in basic research, and the relationship of their interests to the overall program. These smmaries are intended to provide a rapid means for becoming acquainted with the Chemicalmore » Sciences program for members of the scientific and technological public, and interested persons in the Legislative and Executive Branches of the Government, in order to indicate the areas of research supported by the Division and energy technologies which may be advanced by use of basic knowledge discovered in this program. Scientific excellence is a major criterion applied in the selection of research supported by Chemical Sciences. Another important consideration is the identifying of chemical, physical and chemical engineering subdisciplines which are advancing in ways which produce new information related to energy, needed data, or new ideas.« less

Plant maintenance and advanced reactors issue, 2004

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

Agnihotri, Newal

2004-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Optimism about the future of nuclear power, by Ruth G. Shaw, Duke Power Company; Licensed in three countries, by GE Energy; Enhancing public acceptance, by Westinghouse Electric Company; Standardized MOV program, by Ted Neckowicz, Exelon; Inservice testing, by Steven Unikewicz, U.S. Nuclear Regulatory Commission; Asian network for education, Fatimah Mohd Amin, Malaysian Institute for Nuclear Technology Research; and, Cooling water intake optimization, by Jeffrey M. Jones and Bert Mayer, P.E., Framatome ANP.

Chemistry Division. Quarterly progress report for period ending June 30, 1949

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

Not Available

1949-09-14

Progress reports are presented for the following tasks: (1) nuclear and chemical properties of heavy elements (solution chemistry, phase rule studies); (2) nuclear and chemical properties of elements in the fission product region; (3) general nuclear chemistry; (4) radio-organic chemistry; (5) chemistry of separations processes; (6) physical chemistry and chemical physics; (7) radiation chemistry; (8) physical measurements and instrumentation; and (9) analytical chemistry. The program of the chemistry division is divided into two efforts of approximately equal weight with respect to number of personnel, chemical research, and analytical service for the Laboratory. The various research problems fall into the followingmore » classifications: (1) chemical separation processes for isolation and recovery of fissionable material, production of radioisotopes, and military applications; (2) reactor development; and (3) fundamental research.« less

Reactor Application for Coaching Newbies

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

2015-06-17

RACCOON is a Moose based reactor physics application designed to engage undergraduate and first-year graduate students. The code contains capabilities to solve the multi group Neutron Diffusion equation in eigenvalue and fixed source form and will soon have a provision to provide simple thermal feedback. These capabilities are sufficient to solve example problems found in Duderstadt & Hamilton (the typical textbook of senior level reactor physics classes). RACCOON does not contain any advanced capabilities as found in YAK.

THE ARMOUR DUST FUELED REACTOR (ADFR)

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

Krucoff, D.

1958-01-01

The A-DFR is based on the use of a fissionable dust carried in a gas. This fuel ferm offers promise of a major economic advance through the use of 2,000 to 3,000 F operating temperatures and a low cost fuel cycle. The development program is described that was initiated to investigate experimentally the proposed fuel and study analytically other reactor characteristics. A brief review of the reactor concept is presented. (W.D.M.)

10 CFR 110.41 - Executive Branch review.

Code of Federal Regulations, 2010 CFR

2010-01-01

... export involving assistance to end uses related to isotope separation, chemical reprocessing, heavy water production, advanced reactors, or the fabrication of nuclear fuel containing plutonium, except for exports of... foreign reactor. (8) An export involving radioactive waste. (9) An export to any country listed in § 110...

Advanced Diesel Oil Fuel Processor Development

DTIC Science & Technology

1986-06-01

water exit 29 sample quencher: gas sample line inlet 30 sample quencher: gas sample line exit 31 sample quencher: cooling water inlet 32 desulfuriser ...exit line 33, 34 desulfurimer 35 heat exchanger: process gas exit (to desulfuriser ) 38 shift reactor inlet (top) 37 shift reactor: cooling air exit

The Dynomak: An advanced spheromak reactor system with imposed-dynamo current drive and next-generation nuclear power technologies

NASA Astrophysics Data System (ADS)

Sutherland, D. A.; Jarboe, T. R.; Marklin, G.; Morgan, K. D.; Nelson, B. A.

2013-10-01

A high-beta spheromak reactor system has been designed with an overnight capital cost that is competitive with conventional power sources. This reactor system utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER developed cryogenic pumping systems were implemented in this design on the basis of technological feasibility. A tritium breeding ratio of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High-temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%. A paper concerning the Dynomak reactor design is currently being reviewed for publication.

Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

NASA Astrophysics Data System (ADS)

Dutta, N. G.

2012-11-01

Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

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

Metal fires and their implications for advanced reactors.

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

Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean

This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in thesemore » areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety analysis capabilities of the advanced-reactor community for directly relevant scenarios. Beyond the focus on the thermally-interacting and smoldering sodium pool fires, experimental and analysis capabilities for sodium spray fires have also been developed in this project.« less

Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

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

Brown, Nicholas R; Wysocki, Aaron J; Terrani, Kurt A

2016-01-01

Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermalmore » hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores due to hardened (or softened) spectrum. This study shows minimal impact of SiC-based cladding configurations on the transient response versus reference zirconium-based cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore the FeCrAl-based cases have a more rapid fuel thermal expansion rate and the resultant pellet-cladding interaction occurs more rapidly.« less

Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

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

D.M. McEligot; K. G. Condie; G. E. McCreery

2005-10-01

Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generationmore » IV program.« less

77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-20

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0010] Knowledge and Abilities Catalog for Nuclear Power... comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant... developed using this Catalog along with the Operator Licensing Examination Standards for Power Reactors...

An eye on reactor and computer control

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

Schryver, J.; Knee, B.

1992-01-01

At ORNL computer software has been developed to make possible an improved eye-gaze measurement technology. Such an inovation could be the basis for advanced eye-gaze systems that may have applications in reactor control, software development, cognitive engineering, evaluation of displays, prediction of mental workloads, and military target recognition.

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...

Key Assets for a Sustainable Low Carbon Energy Future

NASA Astrophysics Data System (ADS)

Carre, Frank

2011-10-01

Since the beginning of the 21st century, concerns of energy security and climate change gave rise to energy policies focused on energy conservation and diversified low-carbon energy sources. Provided lessons of Fukushima accident are evidently accounted for, nuclear energy will probably be confirmed in most of today's nuclear countries as a low carbon energy source needed to limit imports of oil and gas and to meet fast growing energy needs. Future challenges of nuclear energy are then in three directions: i) enhancing safety performance so as to preclude any long term impact of severe accident outside the site of the plant, even in case of hypothetical external events, ii) full use of Uranium and minimization long lived radioactive waste burden for sustainability, and iii) extension to non-electricity energy products for maximizing the share of low carbon energy source in transportation fuels, industrial process heat and district heating. Advanced LWRs (Gen-III) are today's best available technologies and can somewhat advance nuclear energy in these three directions. However, breakthroughs in sustainability call for fast neutron reactors and closed fuel cycles, and non-electric applications prompt a revival of interest in high temperature reactors for exceeding cogeneration performances achievable with LWRs. Both types of Gen-IV nuclear systems by nature call for technology breakthroughs to surpass LWRs capabilities. Current resumption in France of research on sodium cooled fast neutron reactors (SFRs) definitely aims at significant progress in safety and economic competitiveness compared to earlier reactors of this type in order to progress towards a new generation of commercially viable sodium cooled fast reactor. Along with advancing a new generation of sodium cooled fast reactor, research and development on alternative fast reactor types such as gas or lead-alloy cooled systems (GFR & LFR) is strategic to overcome technical difficulties and/or political opposition specific to sodium. In conclusion, research and technology breakthroughs in nuclear power are needed for shaping a sustainable low carbon future. International cooperation is key for sharing costs of research and development of the required novel technologies and cost of first experimental reactors needed to demonstrate enabling technologies. At the same time technology breakthroughs are developed, pre-normative research is required to support codification work and harmonized regulations that will ultimately apply to safety and security features of resulting innovative reactor types and fuel cycles.

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

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

Belles, Randy; Jain, Prashant K.; Powers, Jeffrey J.

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments usingmore » equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.« less

Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term – Trial Calculation

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

Grabaskas, David; Bucknor, Matthew; Jerden, James

2016-10-01

The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is notmore » without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.« less

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

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos supported the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. Baseline calculations of the PIUS design were performed for active and passive reactor scrams using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following active-system scrams.

Functional requirements of computer systems for the U.S. Geological Survey, Water Resources Division, 1988-97

USGS Publications Warehouse

Hathaway, R.M.; McNellis, J.M.

1989-01-01

Investigating the occurrence, quantity, quality, distribution, and movement of the Nation 's water resources is the principal mission of the U.S. Geological Survey 's Water Resources Division. Reports of these investigations are published and available to the public. To accomplish this mission, the Division requires substantial computer technology to process, store, and analyze data from more than 57,000 hydrologic sites. The Division 's computer resources are organized through the Distributed Information System Program Office that manages the nationwide network of computers. The contract that provides the major computer components for the Water Resources Division 's Distributed information System expires in 1991. Five work groups were organized to collect the information needed to procure a new generation of computer systems for the U. S. Geological Survey, Water Resources Division. Each group was assigned a major Division activity and asked to describe its functional requirements of computer systems for the next decade. The work groups and major activities are: (1) hydrologic information; (2) hydrologic applications; (3) geographic information systems; (4) reports and electronic publishing; and (5) administrative. The work groups identified 42 functions and described their functional requirements for 1988, 1992, and 1997. A few new functions such as Decision Support Systems and Executive Information Systems, were identified, but most are the same as performed today. Although the number of functions will remain about the same, steady growth in the size, complexity, and frequency of many functions is predicted for the next decade. No compensating increase in the Division 's staff is anticipated during this period. To handle the increased workload and perform these functions, new approaches will be developed that use advanced computer technology. The advanced technology is required in a unified, tightly coupled system that will support all functions simultaneously. The new approaches and expanded use of computers will require substantial increases in the quantity and sophistication of the Division 's computer resources. The requirements presented in this report will be used to develop technical specifications that describe the computer resources needed during the 1990's. (USGS)

Nonlocal Effects of Crack Curving.

DTIC Science & Technology

1982-07-01

Structures Research Division Marine Corps Development Langley Research Center and Education Comand Langley Station Quantice , Virginia 22134i~l EaptOn, Vtiina...Advanced lomn 3C128 Research and Technology The Pentagon Washington. D.C. 2054 Washington, D.C* 20301 Air Force Dr. G Sans National Science Foundation... Science and Technology Division (FIS) Washingtono DOC* 20540 AM (XIM)Director Chief Applied Mechanics Group Defense Nuclear Agency U.S.. Air Force

Advances in nickel hydrogen technology at Yardney Battery Division

NASA Technical Reports Server (NTRS)

Bentley, J. G.; Hall, A. M.

1987-01-01

The current major activites in nickel hydrogen technology being addressed at Yardney Battery Division are outlined. Five basic topics are covered: an update on life cycle testing of ManTech 50 AH NiH2 cells in the LEO regime; an overview of the Air Force/industry briefing; nickel electrode process upgrading; 4.5 inch cell development; and bipolar NiH2 battery development.

Advanced Aircraft Electrical System Control Technology Demonstrator. Phase I. Requirements Analysis and Conceptual Design.

DTIC Science & Technology

1981-07-01

System 13 (7) Flight Critical Power 15 (8) Power Bus Configuration 16 b. System Control and Protection 20...includes the main buses, external power receptacles and distribution feeders. The function of the distribution protection system * is mainly to provide...TechnicaI rea Manager Power Systems Branch Power Systems B nch Aerospace Power Division Aerospace Power Division FOR .AKE D . REAMS Chief,

Nb3Sn Superconductor Loss Study

DTIC Science & Technology

1988-01-08

ABERO PROPULSION LABORATORY AIR FORCE WRIGHT AERONAUTICAL LABORATORIES AIR FORCE SYSTEMS COMMAND WRIGHT-PATTERSON AIR FORCE BASE, OHIO 45433-6563 93...Advanced Power Systems Branch Aerospace Power Division Aero Propulsion & Power Laboratory FOR THE COMMANDER MiCHAEL D. BRAYDICH, Maj, USAF Deputy Director...Aerospace Power Division Aero Propulsion & Power Laboratory IF YOUR ADDRESS HAS CHANGED, IF YOU WISH TO BE REMOVED FROM OUR MAIUNG LIST, OR IF THE

A Fundamental Study of the Electromagnetic Properties of Advanced Composite Materials

DTIC Science & Technology

1978-07-01

MKDC), Space and Missile Systems Organization (SAMSO). Aeronautical System Division (ASD), Electronic Systems Division ( ESD ), Air Force Avionics...discussions, the work reported involved only one fiber type--Thornel T300 as used in Narmco 5208 pre-preg tapes . Individual graphite fibers have radii... teflon coated tweezers to separate individual fibers from the bundle. Microscopic observation and a steady hand during this procedure improved the

Work of the Bureau of Education for the Natives of Alaska. Bulletin, 1925, No. 16

ERIC Educational Resources Information Center

Hamilton, William

1925-01-01

This bulletin documents the work of the Alaska division of the Bureau of Education. It reports advance sheets for the biennial survey of education, 1922-1924. The Alaska division is required to make provision for the education of the natives of Alaska, extend to them all possible medical relief, train them to self-support, and, so far as possible,…

Website for the Space Science Division

NASA Technical Reports Server (NTRS)

Schilling, James; DeVincenzi, Donald (Technical Monitor)

2002-01-01

The Space Science Division at NASA Ames Research Center is dedicated to research in astrophysics, exobiology, advanced life support technologies, and planetary science. These research programs are structured around Astrobiology (the study of life in the universe and the chemical and physical forces and adaptions that influence life's origin, evolution, and destiny), and address some of the most fundamental questions pursued by science. These questions examine the origin of life and our place in the universe. Ames is recognized as a world leader in Astrobiology. In pursuing our mission in Astrobiology, Space Science Division scientists perform pioneering basic research and technology development.

The Simulator Development for RDE Reactor

NASA Astrophysics Data System (ADS)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

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

DTIC Science & Technology

1980-12-01

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

Conceptual assessment tool for advanced undergraduate electrodynamics

NASA Astrophysics Data System (ADS)

Baily, Charles; Ryan, Qing X.; Astolfi, Cecilia; Pollock, Steven J.

2017-12-01

As part of ongoing investigations into student learning in advanced undergraduate courses, we have developed a conceptual assessment tool for upper-division electrodynamics (E&M II): the Colorado UppeR-division ElectrodyNamics Test (CURrENT). This is a free response, postinstruction diagnostic with 6 multipart questions, an optional 3-question preinstruction test, and accompanying grading rubrics. The instrument's development was guided by faculty-consensus learning goals and research into common student difficulties. It can be used to gauge the effectiveness of transformed pedagogy, and to gain insights into student thinking in the covered topic areas. We present baseline data representing 500 students across 9 institutions, along with validity, reliability, and discrimination measures of the instrument and scoring rubric.

Post-treatment of reclaimed waste water based on an electrochemical advanced oxidation process

NASA Technical Reports Server (NTRS)

Verostko, Charles E.; Murphy, Oliver J.; Hitchens, G. D.; Salinas, Carlos E.; Rogers, Tom D.

1992-01-01

The purification of reclaimed water is essential to water reclamation technology life-support systems in lunar/Mars habitats. An electrochemical UV reactor is being developed which generates oxidants, operates at low temperatures, and requires no chemical expendables. The reactor is the basis for an advanced oxidation process in which electrochemically generated ozone and hydrogen peroxide are used in combination with ultraviolet light irradiation to produce hydroxyl radicals. Results from this process are presented which demonstrate concept feasibility for removal of organic impurities and disinfection of water for potable and hygiene reuse. Power, size requirements, Faradaic efficiency, and process reaction kinetics are discussed. At the completion of this development effort the reactor system will be installed in JSC's regenerative water recovery test facility for evaluation to compare this technique with other candidate processes.

A Methodology for the Development of a Reliability Database for an Advanced Reactor Probabilistic Risk Assessment

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

Grabaskas, Dave; Brunett, Acacia J.; Bucknor, Matthew

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of a reliability database (RDB) methodology to determine applicable reliability data for inclusion in the quantification of the PRA. The RDBmore » method developed during this project seeks to satisfy the requirements of the Data Analysis element of the ASME/ANS Non-LWR PRA standard. The RDB methodology utilizes a relevancy test to examine reliability data and determine whether it is appropriate to include as part of the reliability database for the PRA. The relevancy test compares three component properties to establish the level of similarity to components examined as part of the PRA. These properties include the component function, the component failure modes, and the environment/boundary conditions of the component. The relevancy test is used to gauge the quality of data found in a variety of sources, such as advanced reactor-specific databases, non-advanced reactor nuclear databases, and non-nuclear databases. The RDB also establishes the integration of expert judgment or separate reliability analysis with past reliability data. This paper provides details on the RDB methodology, and includes an example application of the RDB methodology for determining the reliability of the intermediate heat exchanger of a sodium fast reactor. The example explores a variety of reliability data sources, and assesses their applicability for the PRA of interest through the use of the relevancy test.« less

Evaluation Metrics Applied to Accident Tolerant Fuels

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

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

2014-10-01

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

Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

NASA Astrophysics Data System (ADS)

Natesan, K.; Li, Meimei; Chopra, O. K.; Majumdar, S.

2009-07-01

Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.

Improving online risk assessment with equipment prognostics and health monitoring

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

Coble, Jamie B.; Liu, Xiaotong; Briere, Chris

The current approach to evaluating the risk of nuclear power plant (NPP) operation relies on static probabilities of component failure, which are based on industry experience with the existing fleet of nominally similar light water reactors (LWRs). As the nuclear industry looks to advanced reactor designs that feature non-light water coolants (e.g., liquid metal, high temperature gas, molten salt), this operating history is not available. Many advanced reactor designs use advanced components, such as electromagnetic pumps, that have not been used in the US commercial nuclear fleet. Given the lack of rich operating experience, we cannot accurately estimate the evolvingmore » probability of failure for basic components to populate the fault trees and event trees that typically comprise probabilistic risk assessment (PRA) models. Online equipment prognostics and health management (PHM) technologies can bridge this gap to estimate the failure probabilities for components under operation. The enhanced risk monitor (ERM) incorporates equipment condition assessment into the existing PRA and risk monitor framework to provide accurate and timely estimates of operational risk.« less

Assessment for advanced fuel cycle options in CANDU

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

Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

2013-07-01

The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a drivermore » fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.« less

PIE on Safety-Tested AGR-1 Compact 5-1-1

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

Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactormore » (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.« less

Fractions division knowledge of elementary school student: The case of Lala

NASA Astrophysics Data System (ADS)

Purnomo, Yoppy Wahyu; Widowati, Chairunnisa; Aziz, Tian Abdul; Pramudiani, Puri

2017-08-01

Division of fractions is often acknowledged by mysterious rule which is not based on conceptual knowledge. The purpose of the study was to explore elementary school student's knowledge of division fractions. For this purpose, a case study was conducted. The participant of the study was Lala (pseudonym) who enrolled at one elementary school in East Jakarta. The data were collected by administering written test and semi-structured interview respectively. The findings of the study indicated that Lala was able to describe strategy of division fractions as inverse of repeated addition flexibly. She also had basic understanding of fractions division concept as equal sharing, but when she was challenged with advance problems, she performed poorly. Lala also encountered difficulty when dealing with dividing fraction by fraction problem in which she interpreted it as subtraction problem. In this case, her procedural knowledge was likely to be more salient than her conceptual knowledge.

Establishing a portfolio of quality-improvement projects in pediatric surgery through advanced improvement leadership systems.

PubMed

Gerrein, Betsy T; Williams, Christina E; Von Allmen, Daniel

2013-01-01

Formal quality-improvement (QI) projects require that participants are educated in QI methods to provide them with the capability to carry out successful, meaningful work. However, orchestrating a portfolio of projects that addresses the strategic mission of the institution requires an extension of basic QI training to provide the division or business unit with the capacity to successfully develop and manage the portfolio. Advanced Improvement Leadership Systems is a program to help units create a meaningful portfolio. This program, used by the Division of Pediatric General and Thoracic Surgery at Cincinnati Children's Hospital Medical Center, helped establish a portfolio of targeted QI projects designed to achieve outstanding outcomes at competitive costs in multiple clinical areas aligned with the institution's strategic goals (improve disease-based outcomes, patient safety, flow, and patient and family experience). These objectives are addressed in an institutional strategic plan built around 5 core areas: Safety, Productivity, Care Coordination and Outcomes, Patient and Family Experience, and Value. By combining the portfolio of QI projects with improvements in the divisional infrastructure, effective improvement efforts were realized throughout the division. In the 9 months following the program, divisional capability resulted in a 16.5% increase (5.7% to 22.2%) of formally trained staff working on 10 QI teams. Concurrently, a leadership team, designed to coordinate projects, remove barriers, and provide technical support, provided the capacity to pursue this ongoing effort. The Advanced Improvement Leadership Systems program increased the Division's efficiency and effectiveness in pursing the QI mission that is integral at our hospital.

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

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

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

1996-12-31

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

A Methodology for Phased Array Radar Threshold Modeling Using the Advanced Propagation Model (APM)

DTIC Science & Technology

2017-10-01

TECHNICAL REPORT 3079 October 2017 A Methodology for Phased Array Radar Threshold Modeling Using the Advanced Propagation Model (APM...Head 55190 Networks Division iii EXECUTIVE SUMMARY This report summarizes the methodology developed to improve the radar threshold modeling...PHASED ARRAY RADAR CONFIGURATION ..................................................................... 1 3. METHODOLOGY

Autonomous Control of Nuclear Power Plants

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

Basher, H.

2003-10-20

A nuclear reactor is a complex system that requires highly sophisticated controllers to ensure that desired performance and safety can be achieved and maintained during its operations. Higher-demanding operational requirements such as reliability, lower environmental impacts, and improved performance under adverse conditions in nuclear power plants, coupled with the complexity and uncertainty of the models, necessitate the use of an increased level of autonomy in the control methods. In the opinion of many researchers, the tasks involved during nuclear reactor design and operation (e.g., design optimization, transient diagnosis, and core reload optimization) involve important human cognition and decisions that maymore » be more easily achieved with intelligent methods such as expert systems, fuzzy logic, neural networks, and genetic algorithms. Many experts in the field of control systems share the idea that a higher degree of autonomy in control of complex systems such as nuclear plants is more easily achievable through the integration of conventional control systems and the intelligent components. Researchers have investigated the feasibility of the integration of fuzzy logic, neural networks, genetic algorithms, and expert systems with the conventional control methods to achieve higher degrees of autonomy in different aspects of reactor operations such as reactor startup, shutdown in emergency situations, fault detection and diagnosis, nuclear reactor alarm processing and diagnosis, and reactor load-following operations, to name a few. With the advancement of new technologies and computing power, it is feasible to automate most of the nuclear reactor control and operation, which will result in increased safety and economical benefits. This study surveys current status, practices, and recent advances made towards developing autonomous control systems for nuclear reactors.« less

From biofilm ecology to reactors: a focused review.

PubMed

Boltz, Joshua P; Smets, Barth F; Rittmann, Bruce E; van Loosdrecht, Mark C M; Morgenroth, Eberhard; Daigger, Glen T

2017-04-01

Biofilms are complex biostructures that appear on all surfaces that are regularly in contact with water. They are structurally complex, dynamic systems with attributes of primordial multicellular organisms and multifaceted ecosystems. The presence of biofilms may have a negative impact on the performance of various systems, but they can also be used beneficially for the treatment of water (defined herein as potable water, municipal and industrial wastewater, fresh/brackish/salt water bodies, groundwater) as well as in water stream-based biological resource recovery systems. This review addresses the following three topics: (1) biofilm ecology, (2) biofilm reactor technology and design, and (3) biofilm modeling. In so doing, it addresses the processes occurring in the biofilm, and how these affect and are affected by the broader biofilm system. The symphonic application of a suite of biological methods has led to significant advances in the understanding of biofilm ecology. New metabolic pathways, such as anaerobic ammonium oxidation (anammox) or complete ammonium oxidation (comammox) were first observed in biofilm reactors. The functions, properties, and constituents of the biofilm extracellular polymeric substance matrix are somewhat known, but their exact composition and role in the microbial conversion kinetics and biochemical transformations are still to be resolved. Biofilm grown microorganisms may contribute to increased metabolism of micro-pollutants. Several types of biofilm reactors have been used for water treatment, with current focus on moving bed biofilm reactors, integrated fixed-film activated sludge, membrane-supported biofilm reactors, and granular sludge processes. The control and/or beneficial use of biofilms in membrane processes is advancing. Biofilm models have become essential tools for fundamental biofilm research and biofilm reactor engineering and design. At the same time, the divergence between biofilm modeling and biofilm reactor modeling approaches is recognized.

Wireless, in-vessel neutron monitor for initial core-loading of advanced breeder reactors

NASA Technical Reports Server (NTRS)

Delorenzo, J. T.; Kennedy, E. J.; Blalock, T. V.; Rochelle, J. M.; Chiles, M. M.; Valentine, K. H.

1981-01-01

An experimental wireless, in-vessel neutron monitor was developed to measure the reactivity of an advanced breeder reactor as the core is loaded for the first time to preclude an accidental critically incident. The environment is liquid sodium at a temperature of approx. 220 C, with negligible gamma or neutron radiation. With ultrasonic transmission of neutron data, no fundamental limitation was observed after tests at 230 C for 2000 h. The neutron sensitivity was approx. 1 count/s-nv, and the potential data transmission rate was approx. 10,000 counts/s.

Improvements to Nuclear Data and Its Uncertainties by Theoretical Modeling

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

Danon, Yaron; Nazarewicz, Witold; Talou, Patrick

2013-02-18

This project addresses three important gaps in existing evaluated nuclear data libraries that represent a significant hindrance against highly advanced modeling and simulation capabilities for the Advanced Fuel Cycle Initiative (AFCI). This project will: Develop advanced theoretical tools to compute prompt fission neutrons and gamma-ray characteristics well beyond average spectra and multiplicity, and produce new evaluated files of U and Pu isotopes, along with some minor actinides; Perform state-of-the-art fission cross-section modeling and calculations using global and microscopic model input parameters, leading to truly predictive fission cross-sections capabilities. Consistent calculations for a suite of Pu isotopes will be performed; Implementmore » innovative data assimilation tools, which will reflect the nuclear data evaluation process much more accurately, and lead to a new generation of uncertainty quantification files. New covariance matrices will be obtained for Pu isotopes and compared to existing ones. The deployment of a fleet of safe and efficient advanced reactors that minimize radiotoxic waste and are proliferation-resistant is a clear and ambitious goal of AFCI. While in the past the design, construction and operation of a reactor were supported through empirical trials, this new phase in nuclear energy production is expected to rely heavily on advanced modeling and simulation capabilities. To be truly successful, a program for advanced simulations of innovative reactors will have to develop advanced multi-physics capabilities, to be run on massively parallel super- computers, and to incorporate adequate and precise underlying physics. And all these areas have to be developed simultaneously to achieve those ambitious goals. Of particular interest are reliable fission cross-section uncertainty estimates (including important correlations) and evaluations of prompt fission neutrons and gamma-ray spectra and uncertainties.« less

Aggregate Size and Architecture Determine Microbial Activity Balance for One-Stage Partial Nitritation and Anammox ▿

PubMed Central

Vlaeminck, Siegfried E.; Terada, Akihiko; Smets, Barth F.; De Clippeleir, Haydée; Schaubroeck, Thomas; Bolca, Selin; Demeestere, Lien; Mast, Jan; Boon, Nico; Carballa, Marta; Verstraete, Willy

2010-01-01

Aerobic ammonium-oxidizing bacteria (AerAOB) and anoxic ammonium-oxidizing bacteria (AnAOB) cooperate in partial nitritation/anammox systems to remove ammonium from wastewater. In this process, large granular microbial aggregates enhance the performance, but little is known about granulation so far. In this study, three suspended-growth oxygen-limited autotrophic nitrification-denitrification (OLAND) reactors with different inoculation and operation (mixing and aeration) conditions, designated reactors A, B, and C, were used. The test objectives were (i) to quantify the AerAOB and AnAOB abundance and the activity balance for the different aggregate sizes and (ii) to relate aggregate morphology, size distribution, and architecture putatively to the inoculation and operation of the three reactors. A nitrite accumulation rate ratio (NARR) was defined as the net aerobic nitrite production rate divided by the anoxic nitrite consumption rate. The smallest reactor A, B, and C aggregates were nitrite sources (NARR, >1.7). Large reactor A and C aggregates were granules capable of autonomous nitrogen removal (NARR, 0.6 to 1.1) with internal AnAOB zones surrounded by an AerAOB rim. Around 50% of the autotrophic space in these granules consisted of AerAOB- and AnAOB-specific extracellular polymeric substances. Large reactor B aggregates were thin film-like nitrite sinks (NARR, <0.5) in which AnAOB were not shielded by an AerAOB layer. Voids and channels occupied 13 to 17% of the anoxic zone of AnAOB-rich aggregates (reactors B and C). The hypothesized granulation pathways include granule replication by division and budding and are driven by growth and/or decay based on species-specific physiology and by hydrodynamic shear and mixing. PMID:19948857

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.

Coordinating Council. Third Meeting: STI Strategic Plans

NASA Technical Reports Server (NTRS)

1990-01-01

The NASA Scientific and Technical Information Program Coordinating Council conducts meetings after which both modified transcripts of presentations and interactive discussions are published. The theme for the November 1990 meeting was 'STI Strategic Plans'. This theme was the focus of recorded discussions by members of the council. The last section of the report presents visuals on strategic goals for the STI Information Division. NASA's vision is to be at the forefront of advancements in aeronautics, space science, and exploration. More specific NASA goals are listed followed by the STI Division mission statement. The Strategic Goals for the STI Division are outlined as follows: Implement effective management strategies, Accomplish rapid deployment of the NASA STI Network, Seek out and develop cooperative partnerships, Establish the STI Program as an integral part of the NASA R&D effort, Enhance the quality of our products and services through a focus on the customer, Build an attitude of quality throughout the enterprise, Expand the existing participant community, Assert a NASA leadership role for STI policy, and Develop a program for information science R&D. The STI division mission statement appears on the document cover as follows 'The mission of the NASA STI Program is to advance aerospace knowledge, contribute to U.S. competitiveness, and become an integral partner in NASA R&D programs to support NASA goals.'

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

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

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

2016-01-01

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

Divisions of geologic time-major chronostratigraphic and geochronologic units

USGS Publications Warehouse

,

2010-01-01

Effective communication in the geosciences requires consistent uses of stratigraphic nomenclature, especially divisions of geologic time. A geologic time scale is composed of standard stratigraphic divisions based on rock sequences and is calibrated in years. Over the years, the development of new dating methods and the refinement of previous methods have stimulated revisions to geologic time scales. Advances in stratigraphy and geochronology require that any time scale be periodically updated. Therefore, Divisions of Geologic Time, which shows the major chronostratigraphic (position) and geochronologic (time) units, is intended to be a dynamic resource that will be modified to include accepted changes of unit names and boundary age estimates. This fact sheet is a modification of USGS Fact Sheet 2007-3015 by the U.S. Geological Survey Geologic Names Committee.

Electro-Fenton as a feasible advanced treatment process to produce reclaimed water.

PubMed

Durán Moreno, A; Frontana-Uribe, B A; Ramírez Zamora, R M

2004-01-01

The feasibility of the electro-Fenton process to generate simultaneously both of the Fenton's reagent species (Fe2+/H2O2), was assessed as a potentially more economical alternative to the classical Fenton's reaction to produce reclaimed water. An air-saturated combined wastewater (mixture of municipal and laboratory effluents) was treated in discontinuous and continuous reactors at pH = 3.5. The discontinuous reactor was a 2 L electrochemical laboratory cell fitted with concentric graphite and iron electrodes. The continuous reactor tests used a pilot treatment system comprising the aforementioned electrochemical cell, two clarifiers and one sand filter. Several tests were carried out at different conditions of reaction time (0-60 min) and electrical current values (0.2-1.0 A) in the discontinuous reactor. The best operating conditions were 60 min and 1 A without filtration of effluents. At these conditions, in discontinuous and continuous reactors with filtration, the COD, turbidity and color removal were 65-74.8%, 77-92.3% and 80-100%, respectively. Fecal and total coliforms, Escherichia coli, Shigella and Salmonella sp. were not detected at the end of the pilot treatment system. Electrogeneration of the Fenton's reagent is also economical; its cost is one-fifth the cost reported for Advanced Primary Treatment.

Testing piezoelectric sensors in a nuclear reactor environment

NASA Astrophysics Data System (ADS)

Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

2017-02-01

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

Eyelid reconstruction using the "Hughes" tarsoconjunctival advancement flap: Long-term outcomes in 122 consecutive cases over a 13-year period.

PubMed

McKelvie, James; Ferguson, Reid; Ng, Stephen G J

2017-08-01

This article evaluates the complications and long term functional and cosmetic outcomes of tarsoconjunctival advancement flaps for repairing a range of lower eyelid defects in a large cohort of consecutive cases. A retrospective series of 122 consecutive cases of eyelid reconstruction using tarsoconjunctival-advancement flaps was conducted at Waikato Hospital, or Hamilton Eye Clinic, New Zealand. All cases of lid reconstruction using tarsoconjunctival-advancement flaps between January 1, 2001 until April 3, 2014 were included for analysis. All patients provided written consent for surgery and the study complied with New Zealand Health and Disability Ethics Committee guidelines and the Declaration of Helsinki. Data were collected on patient demographics, lesion histology, defect size, adjuvant surgical procedures required for reconstruction, surgical and postoperative complications, cosmesis and patient satisfaction. Patients requiring lower eyelid reconstruction were predominantly male (56%) and basal cell carcinoma was the most common pathology (>80%). Male gender was associated with larger tarsoconjunctival-advancement-flap width (P-value = 0.0432), larger maximum flap width (20 vs 15 mm), and required on average more adjuvant procedures for reconstruction (1.80 vs 1.48, P-value = 0.02). Mean duration to flap division was 37 days and decreased over the duration of the study. Complicated cases were associated with shorter duration to flap division. Mean follow-up was 7 months, complications were observed in 14% with revision required in 4%. Tarsoconjunctival flap reconstruction of the lower lid is suitable for a range of defect sizes and produces excellent functional and cosmetic outcomes. Complications are relatively infrequent and may be associated in some cases with decreased duration to flap division.

Transient Load Following and Control Analysis of Advanced S-CO2 Power Conversion with Dry Air Cooling

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

Moisseytsev, Anton; Sienicki, James J.

2016-01-01

Supercritical carbon dioxide (S-CO2) Brayton cycles are under development as advanced energy converters for advanced nuclear reactors, especially the Sodium-Cooled Fast Reactor (SFR). The use of dry air cooling for direct heat rejection to the atmosphere ultimate heat sink is increasingly becoming a requirement in many regions due to restrictions on water use. The transient load following and control behavior of an SFR with an S-CO2 cycle power converter utilizing dry air cooling have been investigated. With extension and adjustment of the previously existing control strategy for direct water cooling, S-CO2 cycle power converters can also be used for loadmore » following operation in regions where dry air cooling is a requirement« less

FY 2017-Influence of Sodium Environment on the Tensile Properties of Advanced Alloys

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

Natesan, K.; Li, Meimei; Chen, Wei-Ying

This report provides an update on the understanding of the effects of sodium exposures on tensile properties of advanced alloy 709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY17 (M3AT-17AN1602093), under the Work Package AT-17AN160209, “Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of Advanced Reactor Technologies Program. Three laboratory-size heats of Alloy 709 austenitic steel were investigated in liquid sodium environments at 550-650°C to understand its corrosion behaviour, microstructural evolution, and tensile properties. In addition, a commercial scale heat has beenmore » produced and hot-rolled into plates.« less

Brookhaven highlights for fiscal year 1991, October 1, 1990--September 30, 1991

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

Rowe, M.S.; Cohen, A.; Greenberg, D.

1991-12-31

This report highlights Brookhaven National Laboratory`s activities for fiscal year 1991. Topics from the four research divisions: Computing and Communications, Instrumentation, Reactors, and Safety and Environmental Protection are presented. The research programs at Brookhaven are diverse, as is reflected by the nine different scientific departments: Accelerator Development, Alternating Gradient Synchrotron, Applied Science, Biology, Chemistry, Medical, National Synchrotron Light Source, Nuclear Energy, and Physics. Administrative and managerial information about Brookhaven are also disclosed. (GHH)

Brookhaven highlights for fiscal year 1991, October 1, 1990--September 30, 1991

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

Rowe, M.S.; Cohen, A.; Greenberg, D.

1991-01-01

This report highlights Brookhaven National Laboratory's activities for fiscal year 1991. Topics from the four research divisions: Computing and Communications, Instrumentation, Reactors, and Safety and Environmental Protection are presented. The research programs at Brookhaven are diverse, as is reflected by the nine different scientific departments: Accelerator Development, Alternating Gradient Synchrotron, Applied Science, Biology, Chemistry, Medical, National Synchrotron Light Source, Nuclear Energy, and Physics. Administrative and managerial information about Brookhaven are also disclosed. (GHH)

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).

High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

2009-01-01

An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

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

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

Wheeler, Timothy A.; Liao, Huafei

2014-12-01

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

Advanced Space Nuclear Reactors from Fiction to Reality

NASA Astrophysics Data System (ADS)

Popa-Simil, L.

The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

Overview of the present progress and activities on the CFETR

NASA Astrophysics Data System (ADS)

Wan, Yuanxi; Li, Jiangang; Liu, Yong; Wang, Xiaolin; Chan, Vincent; Chen, Changan; Duan, Xuru; Fu, Peng; Gao, Xiang; Feng, Kaiming; Liu, Songlin; Song, Yuntao; Weng, Peide; Wan, Baonian; Wan, Farong; Wang, Heyi; Wu, Songtao; Ye, Minyou; Yang, Qingwei; Zheng, Guoyao; Zhuang, Ge; Li, Qiang; CFETR Team

2017-10-01

The China Fusion Engineering Test Reactor (CFETR) is the next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). CFETR will be operated in two phases. Steady-state operation and self-sufficiency will be the two key issues for Phase I with a modest fusion power of up to 200 MW. Phase II aims for DEMO validation with a fusion power over 1 GW. Advanced H-mode physics, high magnetic fields up to 7 T, high frequency electron cyclotron resonance heating and lower hybrid current drive together with off-axis negative-ion neutral beam injection will be developed for achieving steady-state advanced operation. The recent detailed design, research and development (R&D) activities including integrated modeling of operation scenarios, high field magnet, material, tritium plant, remote handling and future plans are introduced in this paper.

The role of inertial fusion energy in the energy marketplace of the 21st century and beyond

NASA Astrophysics Data System (ADS)

John Perkins, L.

The viability of inertial fusion in the 21st century and beyond will be determined by its ultimate cost, complexity, and development path relative to other competing, long term, primary energy sources. We examine this potential marketplace in terms of projections for population growth, energy demands, competing fuel sources and environmental constraints (CO 2), and show that the two competitors for inertial fusion energy (IFE) in the medium and long term are methane gas hydrates and advanced, breeder fission; both have potential fuel reserves that will last for thousands of years. Relative to other classes of fusion concepts, we argue that the single largest advantage of the inertial route is the perception by future customers that the IFE fusion power core could achieve credible capacity factors, a result of its relative simplicity, the decoupling of the driver and reactor chamber, and the potential to employ thick liquid walls. In particular, we show that the size, cost and complexity of the IFE reactor chamber is little different to a fission reactor vessel of the same thermal power. Therefore, relative to fission, because of IFE's tangible advantages in safety, environment, waste disposal, fuel supply and proliferation, our research in advanced targets and innovative drivers can lead to a certain, reduced-size driver at which future utility executives will be indifferent to the choice of an advanced fission plant or an advanced IFE power plant; from this point on, we have a competitive commercial product. Finally, given that the major potential customer for energy in the next century is the present developing world, we put the case for future IFE "reservations" which could be viable propositions providing sufficient reliability and redundancy can be realized for each modular reactor unit.

Development/Modernization of an Advanced Non-Light Water Reactor Probabilistic Risk Assessment

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

Henneke, Dennis W.; Robinson, James

In 2015, GE Hitachi Nuclear Energy (GEH) teamed with Argonne National Laboratory (Argonne) to perform Research and Development (R&D) of next-generation Probabilistic Risk Assessment (PRA) methodologies for the modernization of an advanced non-Light Water Reactor (non-LWR) PRA. This effort built upon a PRA developed in the early 1990s for GEH’s Power Reactor Inherently Safe Module (PRISM) Sodium Fast Reactor (SFR). The work had four main tasks: internal events development modeling the risk from the reactor for hazards occurring at-power internal to the plant; an all hazards scoping review to analyze the risk at a high level from external hazards suchmore » as earthquakes and high winds; an all modes scoping review to understand the risk at a high level from operating modes other than at-power; and risk insights to integrate the results from each of the three phases above. To achieve these objectives, GEH and Argonne used and adapted proven PRA methodologies and techniques to build a modern non-LWR all hazards/all modes PRA. The teams also advanced non-LWR PRA methodologies, which is an important outcome from this work. This report summarizes the project outcomes in two major phases. The first phase presents the methodologies developed for non-LWR PRAs. The methodologies are grouped by scope, from Internal Events At-Power (IEAP) to hazards analysis to modes analysis. The second phase presents details of the PRISM PRA model which was developed as a validation of the non-LWR methodologies. The PRISM PRA was performed in detail for IEAP, and at a broader level for hazards and modes. In addition to contributing methodologies, this project developed risk insights applicable to non-LWR PRA, including focus-areas for future R&D, and conclusions about the PRISM design.« less

High Strength Steel Welding Research

DTIC Science & Technology

2005-05-27

E.S.K. Menon, and M.G. Hall, "Sympathetic Nucleation: An Overview," Materials Science and Engineering B, Solid State Materials for Advanced Technology ...GTAW of Titanium Using Flux-cored Wire with Magnesium Fluoride Kook-soo Bang’, Greg Chirieleison 2, and Stephen Liu 2 1 Division of Advanced Materials...O CHAPTER I O INTRODUCTION * The application of advanced high strength low alloy (HSLA) steels has been * limited by the availability of suitable

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

Control console replacement at the WPI Reactor

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

Not Available

1992-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Positron Annihilation Spectroscopy Characterization of Nanostructural Features in Reactor Steels

NASA Astrophysics Data System (ADS)

Glade, Stephen; Wirth, Brian; Asoka-Kumar, Palakkal; Sterne, Philip; Alinger, Matthew; Odette, George

2004-03-01

Irradiation embrittlement in nuclear reactor pressure vessel steels results from the formation of a high number density of nanometer sized copper rich precipitates and sub-nanometer defect-solute clusters. We present results of study to characterize the size and compositions of simple binary and ternary Fe-Cu-Mn model alloys and more representative Fe-Cu-Mn-Ni-Si-Mo-C reactor pressure vessel steels using positron annihilation spectroscopy (PAS). Using a recently developed spin-polarized PAS technique, we have also measured the magnetic properties of the nanometer-sized copper rich precipitates. Mn retards the precipitation kinetics and inhibits large vacancy cluster formation, suggesting a strong Mn-vacancy interaction which reduces radiation enhanced diffusion. The spin-polarized PAS measurements reveal the non-magnetic nature of the copper precipitates, discounting the notion that the precipitates contain significant quantities of Fe and providing an upper limit of at most a few percent Fe in the precipitates. PAS results on oxide dispersion-strengthened steel for use in fusion reactors will also be presented. Part of this work was performed under the auspices of the US Department of Energy by the University of California, Lawrence Livermore National Laboratory, under contract No. W-7405-ENG-48 with partial support provided from Basic Energy Sciences, Division of Materials Science.

19 CFR 181.92 - Definitions and general NAFTA advance ruling practice.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 19 Customs Duties 2 2014-04-01 2014-04-01 false Definitions and general NAFTA advance ruling practice. 181.92 Section 181.92 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND... National Commodity Specialist Division or by such other office as designated by the Commissioner of Customs...

19 CFR 181.92 - Definitions and general NAFTA advance ruling practice.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 19 Customs Duties 2 2013-04-01 2013-04-01 false Definitions and general NAFTA advance ruling practice. 181.92 Section 181.92 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND... National Commodity Specialist Division or by such other office as designated by the Commissioner of Customs...

77 FR 8209 - Quality Assurance Requirements for Continuous Opacity Monitoring Systems at Stationary Sources

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-14

... Transfer and Advancement Act Section 12(d) of the National Technology Transfer and Advancement Act of 1995... Division, Measurement Technology Group (Mail Code: E143-02), Research Triangle Park, NC 27711; telephone... significant economic impact on a substantial number of small entities. Small entities include small businesses...

Genomic Advances to Improve Biomass for Biofuels (LBNL Science at the Theater)

ScienceCinema

Rokhsar, Daniel [USDOE Joint Genome Institute (JGI), Walnut Creek, CA (United States)

2018-05-24

Lawrence Berkeley National Lab bioscientist Daniel Rokhsar discusses genomic advances to improve biomass for biofuels. He presented his talk Feb. 11, 2008 in Berkeley, California as part of Berkeley Lab's community lecture series. Rokhsar works with the U.S. Department of Energy's Joint Genome Institute and Berkeley Lab's Genomics Division.

Advances in the understanding of dairy and cheese flavors: Symposium Introduction

USDA-ARS?s Scientific Manuscript database

A symposium titled “Advances in the Understanding of Dairy and Cheese Flavors” was held in September 2013 at the American Chemical Society’s 246th National Meeting in Indianapolis, IN. The symposium, which was sponsored by the Division of Agricultural and Food Chemistry, was to discuss the state of...

Advanced Very High Resolution Radiometer - AVHRR - NOAA Satellite

Science.gov Websites

Information System (NOAASIS); Office of Satellite and Product Operations » DOC » NOAA  » NESDIS » NOAASIS NOAA Satellite Information System Advanced Very High Resolution Radiometer - AVHRR The ) or the USGS AVHRR site. Satellite Products and Services Division Direct Services Branch Phone: 301

Genomic Advances to Improve Biomass for Biofuels (Genomics and Bioenergy)

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

Rokhsar, Daniel

2008-02-11

Lawrence Berkeley National Lab bioscientist Daniel Rokhsar discusses genomic advances to improve biomass for biofuels. He presented his talk Feb. 11, 2008 in Berkeley, California as part of Berkeley Lab's community lecture series. Rokhsar works with the U.S. Department of Energy's Joint Genome Institute and Berkeley Lab's Genomics Division.

Genomic Advances to Improve Biomass for Biofuels (LBNL Science at the Theater)

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

Rokhsar, Daniel

2008-02-11

Lawrence Berkeley National Lab bioscientist Daniel Rokhsar discusses genomic advances to improve biomass for biofuels. He presented his talk Feb. 11, 2008 in Berkeley, California as part of Berkeley Lab's community lecture series. Rokhsar works with the U.S. Department of Energy's Joint Genome Institute and Berkeley Lab's Genomics Division.

CFD Study of Full-Scale Aerobic Bioreactors: Evaluation of Dynamic O2 Distribution, Gas-Liquid Mass Transfer and Reaction

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

Humbird, David; Sitaraman, Hariswaran; Stickel, Jonathan

If advanced biofuels are to measurably displace fossil fuels in the near term, they will have to operate at levels of scale, efficiency, and margin unprecedented in the current biotech industry. For aerobically-grown products in particular, scale-up is complex and the practical size, cost, and operability of extremely large reactors is not well understood. Put simply, the problem of how to attain fuel-class production scales comes down to cost-effective delivery of oxygen at high mass transfer rates and low capital and operating costs. To that end, very large reactor vessels (>500 m3) are proposed in order to achieve favorable economiesmore » of scale. Additionally, techno-economic evaluation indicates that bubble-column reactors are more cost-effective than stirred-tank reactors in many low-viscosity cultures. In order to advance the design of extremely large aerobic bioreactors, we have performed computational fluid dynamics (CFD) simulations of bubble-column reactors. A multiphase Euler-Euler model is used to explicitly account for the spatial distribution of air (i.e., gas bubbles) in the reactor. Expanding on the existing bioreactor CFD literature (typically focused on the hydrodynamics of bubbly flows), our simulations include interphase mass transfer of oxygen and a simple phenomenological reaction representing the uptake and consumption of dissolved oxygen by submerged cells. The simulations reproduce the expected flow profiles, with net upward flow in the center of column and downward flow near the wall. At high simulated oxygen uptake rates (OUR), oxygen-depleted regions can be observed in the reactor. By increasing the gas flow to enhance mixing and eliminate depleted areas, a maximum oxygen transfer (OTR) rate is obtained as a function of superficial velocity. These insights regarding minimum superficial velocity and maximum reactor size are incorporated into NREL's larger techno-economic models to supplement standard reactor design equations.« less

A KINETIC MODEL FOR H2O2/UV PROCESS IN A COMPLETELY MIXED BATCH REACTOR. (R825370C076)

EPA Science Inventory

A dynamic kinetic model for the advanced oxidation process (AOP) using hydrogen peroxide and ultraviolet irradiation (H₂O₂/UV) in a completely mixed batch reactor (CMBR) is developed. The model includes the known elementary chemical and photochemical reac...

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

Logan, B.G.

A recently completed two-year study of a commercial tandem mirror reactor design (Mirror Advanced Reactor Study (MARS)) is briefly reviewed. The end plugs are designed for trapped particle stability, MHD ballooning, balanced geodesic curvature, and small radial electric fields in the central cell. New technologies such as lithium-lead blankets, 24T hybrid coils, gridless direct converters and plasma halo vacuum pumps are highlighted.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-30

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-30

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

A unique nuclear thermal rocket engine using a particle bed reactor

NASA Astrophysics Data System (ADS)

Culver, Donald W.; Dahl, Wayne B.; McIlwain, Melvin C.

1992-01-01

Aerojet Propulsion Division (APD) studied 75-klb thrust Nuclear Thermal Rocket Engines (NTRE) with particle bed reactors (PBR) for application to NASA's manned Mars mission and prepared a conceptual design description of a unique engine that best satisfied mission-defined propulsion requirements and customer criteria. This paper describes the selection of a sprint-type Mars transfer mission and its impact on propulsion system design and operation. It shows how our NTRE concept was developed from this information. The resulting, unusual engine design is short, lightweight, and capable of high specific impulse operation, all factors that decrease Earth to orbit launch costs. Many unusual features of the NTRE are discussed, including nozzle area ratio variation and nozzle closure for closed loop after cooling. Mission performance calculations reveal that other well known engine options do not support this mission.

Personnel Attrition Rates in Historical Land Combat Operations: An Annotated Bibliography

DTIC Science & Technology

1993-06-01

1986, Combat Casualties Among US Navy Personnel in Vietnam. Cochrane-1959, The 3rd Division at Chateau Thierry , July 1918. Cockrell-1974, Research Study...Chateau Thierry , July 1918. CIVIL Fox-1889, Regimental Losses in the American Civil War, 1861-1865. Livermore-1900, Numbers and Losses in the Civil War... Thierry , July 1918. Wainstein-1973b, Rates of Advance in Infantry Division Attacks in the Normandy-Northern France and Siegfried Line Campaigns. DURATION

Remarks of the Assistant Attorney General, Civil Rights Division, before the Paralyzed Veterans of America Disability Rights Conference (Washington, D.C., April 5, 1984).

ERIC Educational Resources Information Center

Reynolds, Wm. Bradford

The commitment of the Assistant Attorney General of the Civil Rights Division, in the U.S. Department of Justice, to the enforcement of the Rehabilitation Act of 1973 is described. Recent federal initiatives under this act to advance the rights of disabled people in terms of employment, the rights of handicapped infants, and the enforcement of…

Best Practices Case Study: Pulte Homes and Communities of Del Webb, Las Vegas Division

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

None

2009-10-01

Case study of Pulte Homes Las Vegas Division, who certified nearly 1,200 homes to the DOE Builders Challenge between 2008 and 2012. All of the homes by Las Vegas’ biggest builder achieved HERS scores of lower than 70, and many floor plans got down to the mid 50s, with ducts located in sealed attics insulated along the roof line, advanced framing, and extra attention to air sealing.

Identifying Strategic Scientific Opportunities

Cancer.gov

As NCI's central scientific strategy office, CRS collaborates with the institute's divisions, offices, and centers to identify research opportunities to advance NCI's vision for the future of cancer research.

Advanced heat exchanger development for molten salts

DOE PAGES

Sabharwall, Piyush; Clark, Denis; Glazoff, Michael; ...

2014-12-01

This study addresses present work concerned with advanced heat exchanger development for molten salt in nuclear and non nuclear thermal systems. The molten salt systems discussed herein use alloys, such as Hastelloy N and 242, which show corrosion resistance to molten salt at nominal operating temperatures up to 700°C. These alloys were diffusion welded, and the corresponding information is presented. Test specimens were prepared for exposing diffusion welds to molten salt environments. Hastelloy N and 242 were found to be weldable by diffusion welding, with ultimate tensile strengths about 90% of base metal values. Both diffusion welds and sheet materialmore » in Hastelloy N were corrosion tested in?58 mol% KF and 42 mol% ZrF4 at 650, 700, and 850°C for 200, 500, and 1,000 hours. Corrosion rates found were similar between welded and nonwelded materials, typically <10 mils per year. For materials of construction, nickel and alloys with dense nickel coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Hence, additional testing of selected alloys for resistance to intergranular corrosion is needed, as is a determination of corrosion rate as a function of contaminant type and alloy composition with respect to chromium and carbon to better define the optimal chromium and carbon composition, independent of galvanic or differential solubility effects. Also presented is the division of the nuclear reactor and high temperature components per ASME standards, along with design requirements for a subcritical Rankine power cycle heat exchanger that has to overcome pressure difference of about 17 MPa.« less

SUPERFUND TREATABILITY CLEARINGHOUSE ...

EPA Pesticide Factsheets

This newsletter reports on the Huber Technology Groups (HTG) high temperature advanced hazardous waste treatment technology capable of very high destruction and removal efficiencies of various hazardous wastes. This newsletter addresses the destruction of PCBs in an EPA certification test of the HTG Advanced Electric Reactor. provide information

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

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

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

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

Structural materials issues for the next generation fission reactors

NASA Astrophysics Data System (ADS)

Chant, I.; Murty, K. L.

2010-09-01

Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

Development of Cross Section Library and Application Programming Interface (API)

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

Lee, C. H.; Marin-Lafleche, A.; Smith, M. A.

2014-04-09

The goal of NEAMS neutronics is to develop a high-fidelity deterministic neutron transport code termed PROTEUS for use on all reactor types of interest, but focused primarily on sodium-cooled fast reactors. While PROTEUS-SN has demonstrated good accuracy for homogeneous fast reactor problems and partially heterogeneous fast reactor problems, the simulation results were not satisfactory when applied on fully heterogeneous thermal problems like the Advanced Test Reactor (ATR). This is mainly attributed to the quality of cross section data for heterogeneous geometries since the conventional cross section generation approach does not work accurately for such irregular and complex geometries. Therefore, onemore » of the NEAMS neutronics tasks since FY12 has been the development of a procedure to generate appropriate cross sections for a heterogeneous geometry core.« less

The ENABLER - Based on proven NERVA technology

NASA Astrophysics Data System (ADS)

Livingston, Julie M.; Pierce, Bill L.

The ENABLER reactor for use in a nuclear thermal propulsion engine uses the technology developed in the NERVA/Rover program, updated to incorporate advances in the technology. Using composite fuel, higher power densities per fuel element, improved radiation resistant control components and the advancements in use of carbon-carbon materials; the ENABLER can provide a specific impulse of 925 seconds, an engine thrust to weight (excluding reactor shield) approaching five, an improved initial mass in low Earth orbit and a consequent reduction in launch costs and logistics problems. This paper describes the 75,000 lbs thrust ENABLER design which is a low cost, low risk approach to meeting tommorrow's space propulsion needs.

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

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

J. Rempe; D. Knudson; J. Daw

2014-01-01

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

A Practical Approach to Starting Fission Surface Power Development

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2006-01-01

The Prometheus Power and Propulsion Program has been reformulated to address NASA needs relative to lunar and Mars exploration. Emphasis has switched from the Jupiter Icy Moons Orbiter (JIMO) flight system development to more generalized technology development addressing Fission Surface Power (FSP) and Nuclear Thermal Propulsion (NTP). Current NASA budget priorities and the deferred mission need date for nuclear systems prohibit a fully funded reactor Flight Development Program. However, a modestly funded Advanced Technology Program can and should be conducted to reduce the risk and cost of future flight systems. A potential roadmap for FSP technology development leading to possible flight applications could include three elements: 1) Conceptual Design Studies, 2) Advanced Component Technology, and 3) Non-Nuclear System Testing. The Conceptual Design Studies would expand on recent NASA and DOE analyses while increasing the depth of study in areas of greatest uncertainty such as reactor integration and human-rated shielding. The Advanced Component Technology element would address the major technology risks through development and testing of reactor fuels, structural materials, primary loop components, shielding, power conversion, heat rejection, and power management and distribution (PMAD). The Non-Nuclear System Testing would provide a modular, technology testbed to investigate and resolve system integration issues.

A Comparative Study of Welded ODS Cladding materials for AFCI/GNEP Applications

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

Indrajit Charit; Megan Frary; Darryl Butt

2011-03-31

This research project involved working on the pressure resistance welding of oxide dispersion strengthened (ODS) alloys which will have a large role to play in advanced nuclear reactors. The project also demonstrated the research collaboration between four universities and one nation laboratory (Idaho National Laboratory) with participation from an industry for developing for ODS alloys. These alloys contain a high number density of very fine oxide particles that can impart high temperature strength and radiation damage resistance suitable for in-core applications in advanced reactors. The conventional fusion welding techniques tend to produce porosity-laden microstructure in the weld region and leadmore » to the agglomeration and non-uniform distribution of the neededoxide particles. That is why two solid state welding methods - pressure resistance welding (PRW) and friction stir welding (FSW) - were chosen to be evaluated in this project. The proposal is expected to support the development of Advanced Burner Reactors (ABR) under the GNEP program (now incorporated in Fuel Cycle R&D program). The outcomes of the concluded research include training of graduate and undergraduate students and get them interested in nuclear related research.« less

ASME Material Challenges for Advanced Reactor Concepts

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

Piyush Sabharwall; Ali Siahpush

2013-07-01

This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at highermore » temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.« less

Groundbreaking Ceremony at the NACA's Plum Brook Station

NASA Image and Video Library

1956-09-21

Addison Rothrock, the National Advisory Committee for Aeronautics’s (NACA) Assistant Director of Research, speaks at the groundbreaking ceremony for the Lewis Flight Propulsion Laboratory’s new test reactor at Plum Brook Station. This dedication event was held almost exactly one year after the NACA announced that it would build its $4.5 million nuclear reactor on 500 acres of the army’s 9000-acre Plum Brook Ordnance Works. The site was located in Sandusky, Ohio, approximately 60 miles west of the NACA Lewis laboratory in Cleveland. Lewis Director Raymond Sharp is seated to the left of Rothrock, Congressman Albert Baumhart and NACA Secretary John Victory are to the right. Many government and local officials were on hand for the press conference and ensuing luncheon. In the wake of World War II the military, the Atomic Energy Commission, and the NACA became interested in the use of atomic energy for propulsion and power. A Nuclear Division was established at NACA Lewis in the early 1950s. The division’s request for a 60-megawatt research reactor was approved in 1955. The semi-remote Plum Brook location was selected over 17 other possible sites. Construction of the Plum Brook Reactor Facility lasted five years. By the time of its first trial runs in 1961 the aircraft nuclear propulsion program had been cancelled. The space age had arrived, however, and the reactor would be used to study materials for a nuclear powered rocket.

Ste12/Fab1 phosphatidylinositol-3-phosphate 5-kinase is required for nitrogen-regulated mitotic commitment and cell size control

PubMed Central

Schauries, Marie; Kaczmarek, Adrian; Franz-Wachtel, Mirita; Du, Wei; Krug, Karsten; Maček, Boris; Petersen, Janni

2017-01-01

Tight coupling of cell growth and cell cycle progression enable cells to adjust their rate of division, and therefore size, to the demands of proliferation in varying nutritional environments. Nutrient stress promotes inhibition of Target Of Rapamycin Complex 1 (TORC1) activity. In fission yeast, reduced TORC1 activity advances mitotic onset and switches growth to a sustained proliferation at reduced cell size. A screen for mutants, that failed to advance mitosis upon nitrogen stress, identified a mutant in the PIKFYVE 1-phosphatidylinositol-3-phosphate 5-kinase fission yeast homolog Ste12. Ste12PIKFYVE deficient mutants were unable to advance the cell cycle to reduce cell size after a nitrogen downshift to poor nitrogen (proline) growth conditions. While it is well established that PI(3,5)P2 signalling is required for autophagy and that Ste12PIKFYVE mutants have enlarged vacuoles (yeast lysosomes), neither a block to autophagy or mutants that independently have enlarged vacuoles had any impact upon nitrogen control of mitotic commitment. The addition of rapamycin to Ste12PIKFYVE deficient mutants reduced cell size at division to suggest that Ste12PIKFYVE possibly functions upstream of TORC1. ste12 mutants display increased Torin1 (TOR inhibitor) sensitivity. However, no major impact on TORC1 or TORC2 activity was observed in the ste12 deficient mutants. In summary, Ste12PIKFYVE is required for nitrogen-stress mediated advancement of mitosis to reduce cell size at division. PMID:28273166

Strange bedfellows: The curious case of STAR and Moata

NASA Astrophysics Data System (ADS)

Smith, A. M.; Levchenko, V. A.; Malone, G.

2013-01-01

The 2 MV tandem accelerator named ‘STAR’ was installed at ANSTO in 2003 and commissioned in 2004. It is used for ion beam analysis (IBA) and for radiocarbon measurements by accelerator mass spectrometry (AMS). Convenient space for the accelerator was found in the same building occupied by the decommissioned Argonaut-class nuclear reactor ‘Moata’; the name derives from the aboriginal word for ‘fire stick’ or ‘gentle fire’, appropriate for a 100 kW research reactor. This reactor operated between 1961 and 1995. In 2007 ANSTO’s Engineering Division assembled a team to dismantle and remove the reactor structure, along with its 12.1 tonnes of graphite reflector. The removal and remediation was completed in November 2010 and has won the team a number of prestigious awards. The entire operation was conducted inside a negatively-pressurised double-walled vinyl tent. An air curtain was positioned around the reactor core. The exhaust air from the tent passed through 2-stage HEPA filters before venting through an external stack. Neither ANSTO staff nor contractors received any significant radiation dose during the operation. Given the sensitivity of STAR for detection of 14C/12C (∼10-16) and the numerous routes for production of 14C in the reactor such as 13C(n, γ)14C, 14N(n, p)14C and 17O(n, α)14C there was the potential to directly contaminate the STAR environment with 14C. Furthermore, there was concern that reactor-14C could find its way from this building into the building where the radiocarbon sample preparation laboratories are located. This necessitated restrictions on staff movement between the buildings. We report on 14C control measurements made during and after the operation. These involved direct measurements on the reactor graphite and concrete bioshield, blank targets that were exposed in the building, swipe samples taken inside the tent and around the building and aerosol samples that were collected inside the building throughout the operation.

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

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

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

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

Validation of the USAWC Student War Gaming Model.

DTIC Science & Technology

1983-05-18

from EREZ clogged with traffic; brigade, and SHARON division Egyptian artillery advance command post, follow harassing; fired on RESHEF brigade along...according to ADAN, "The advance of the Egyptian’s 25th Independent Tank Brigade along the Lexicon axis was incredibly slow; they seemed to be doing...in advance of play which artillery units would be played in each of the two legitimate modes described above. The model unit data base was then

A Comparison of Brayton and Stirling Space Nuclear Power Systems for Power Levels from 1 Kilowatt to 10 Megawatts

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2000-01-01

An analytical study was conducted to assess the performance and mass of Brayton and Stirling nuclear power systems for a wide range of future NASA space exploration missions. The power levels and design concepts were based on three different mission classes. Isotope systems, with power levels from 1 to 10 kW, were considered for planetary surface rovers and robotic science. Reactor power systems for planetary surface outposts and bases were evaluated from 10 to 500 kW. Finally, reactor power systems in the range from 100 kW to 10 mW were assessed for advanced propulsion applications. The analysis also examined the effect of advanced component technology on system performance. The advanced technologies included high temperature materials, lightweight radiators, and high voltage power management and distribution.

Eugene P. Wigner's Visionary Contributions to Generations-I through IV Fission Reactors

NASA Astrophysics Data System (ADS)

Carré, Frank

2014-09-01

Among Europe's greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

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

Mendez Cruz, Carmen Margarita; Rochau, Gary E.; Middleton, Bobby

Sandia National Laboratories and General Atomics are pleased to respond to the Advanced Research Projects Agency-Energy (ARPA-e)’s request for information on innovative developments that may overcome various current reactor-technology limitations. The RFI is particularly interested in innovations that enable ultra-safe and secure modular nuclear energy systems. Our response addresses the specific features for reactor designs called out in the RFI, including a brief assessment of the current state of the technologies that would enable each feature and the methods by which they could be best incorporated into a reactor design.

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

The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

DOE PAGES

Ward, Andrew; Downar, Thomas J.; Xu, Y.; ...

2015-04-22

The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, themore » capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.« less

Coupled reactors analysis: New needs and advances using Monte Carlo methodology

DOE PAGES

Aufiero, M.; Palmiotti, G.; Salvatores, M.; ...

2016-08-20

Coupled reactors and the coupling features of large or heterogeneous core reactors can be investigated with the Avery theory that allows a physics understanding of the main features of these systems. However, the complex geometries that are often encountered in association with coupled reactors, require a detailed geometry description that can be easily provided by modern Monte Carlo (MC) codes. This implies a MC calculation of the coupling parameters defined by Avery and of the sensitivity coefficients that allow further detailed physics analysis. The results presented in this paper show that the MC code SERPENT has been successfully modifed tomore » meet the required capabilities.« less

Cure kinetics, morphologies, and mechanical properties of thermoplastic/MWCNT modified multifunctional glassy epoxies prepared via continuous reaction methods

NASA Astrophysics Data System (ADS)

Cheng, Xiaole

The primary goal of this dissertation is to develop a novel continuous reactor method to prepare partially cured epoxy prepolymers for aerospace prepreg applications with the aim of replacing traditional batch reactors. Compared to batch reactors, the continuous reactor is capable of solubilizing and dispersing a broad range of additives including thermoplastic tougheners, stabilizers, nanoparticles and curatives and advancing epoxy molecular weights and viscosities while reducing energy consumption. In order to prove this concept, polyethersulfone (PES) modified 4, 4'-diaminodiphenylsulfone (44DDS)/tetraglycidyl-4, 4'-diaminodiphenylmethane (TGDDM) epoxy prepolymers were firstly prepared using both continuous reactor and batch reactor methods. Kinetic studies confirmed the chain extension reaction in the continuous reactor is similar to the batch reactor, and the molecular weights and viscosities of prepolymers were readily controlled through reaction kinetics. Atomic force microscopy (AFM) confirmed similar cured network morphologies for formulations prepared from batch and continuous reactors. Additionally tensile strength, tensile modulus and fracture toughness analyses concluded mechanical properties of cured epoxy matrices produced from both reactors were equivalent. Effects of multifunctional epoxy compositions on thermoplastics phase-separated morphologies were systematically studied using a combination of AFM with nanomechanical mapping, spectroscopic and calorimetric techniques to provide new insights to tailor cured reaction induced phase separation (CRIPS) in multifunctional epoxy blend networks. Furthermore, how resultant crosslinked glassy polymer network and phase-separated morphologies correlated with mechanical properties are discussed in detail. Multiwall carbon nanotube (MWCNT)/TGDDM epoxy prepolymers were further prepared by combining the successful strategies for advancing epoxy chemistries and dispersing nanotubes using the continuous reactor. Optical microscopy (OM) and scanning electron microscopy (SEM) were used to characterize the MWCNT dispersion states and stabilization in epoxy prepolymer matrix after continuous process and during curing cycles. Additionally, electrical conductivities and mechanical properties of final cured MWCNT/TGDDM composites were measured and discussed in view of their corresponding MWCNT dispersion states. Ternary blends of MWCNT reinforced thermoplastic/epoxy prepolymers were prepared by the continuous reactor. Influence of MWCNT on the CRIPS mechanism and the cured morphologies were systematically investigated using SEM and rheological analysis. Incorporation of MWCNT in thermoplastic/epoxy matrices can lead to a morphological transformation from phase inverted, to co-continuous, and to droplet dispersed morphology. In additional, dynamic mechanical analysis revealed the heterogeneity of MWCNT dispersion in thermoplastic/thermosets systems.

SUPERFUND TREATABILITY CLEARINGHOUSE: ADVANCED ELECTRIC REACTOR (AER) FOR THE TREATMENT OF DIOXIN- CONTAMINATED SOILS

EPA Science Inventory

This newsletter reports on the Huber Technology Groups (HTG) high temperature advanced hazardous waste treatment technology capable of very high destruction and removal efficiencies of various hazardous wastes. This newsletter addresses the destruction of PCBs in an EPA certifi...

Issues relating to spent nuclear fuel storage on the Oak Ridge Reservation

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

Klein, J.A.; Turner, D.W.

1994-12-31

Currently, about 2,800 metric tons of spent nuclear fuel (SNF) is stored in the US, 1,000 kg of SNF (or about 0.03% of the nation`s total) are stored at the US Department of Energy (DOE) complex in Oak Ridge, Tennessee. However small the total quantity of material stored at Oak Ridge, some of the material is quite singular in character and, thus, poses unique management concerns. The various types of SNF stored at Oak Ridge will be discussed including: (1) High-Flux Isotope Reactor (HFIR) and future Advanced Neutron Source (ANS) fuels; (2) Material Testing Reactor (MTR) fuels, including Bulk Shieldingmore » Reactor (BSR) and Oak Ridge Research Reactor (ORR) fuels; (3) Molten Salt Reactor Experiment (MSRE) fuel; (4) Homogeneous Reactor Experiment (HRE) fuel; (5) Miscellaneous SNF stored in Oak Ridge National Laboratory`s (ORNL`s) Solid Waste Storage Areas (SWSAs); (6) SNF stored in the Y-12 Plant 9720-5 Warehouse including Health. Physics Reactor (HPRR), Space Nuclear Auxiliary Power (SNAP-) 10A, and DOE Demonstration Reactor fuels.« less

Forum: Communication Activism Pedagogy. Communication Activism Pedagogy and Research: Communication Education Scholarship to Promote Social Justice

ERIC Educational Resources Information Center

Frey, Lawrence R.; Palmer, David L.

2017-01-01

The recent formation of the National Communication Association's Activism and Social Justice Division puts a spotlight on the extent to which instructional communication and instructional communication research have advanced--or even should advance--the goals of social justice. To examine this issue, two of the leading scholars on this topic,…

Experiment Needs and Facilities Study Appendix A Transient Reactor Test Facility (TREAT) Upgrade

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

None

The TREAT Upgrade effort is designed to provide significant new capabilities to satisfy experiment requirements associated with key LMFBR Safety Issues. The upgrade consists of reactor-core modifications to supply the physics performance needed for the new experiments, an Advanced TREAT loop with size and thermal-hydraulics capabilities needed for the experiments, associated interface equipment for loop operations and handling, and facility modifications necessary to accommodate operations with the Loop. The costs and schedules of the tasks to be accomplished under the TREAT Upgrade project are summarized. Cost, including contingency, is about 10 million dollars (1976 dollars). A schedule for execution ofmore » 36 months has been established to provide the new capabilities in order to provide timely support of the LMFBR national effort. A key requirement for the facility modifications is that the reactor availability will not be interrupted for more than 12 weeks during the upgrade. The Advanced TREAT loop is the prototype for the STF small-bundle package loop. Modified TREAT fuel elements contain segments of graphite-matrix fuel with graded uranium loadings similar to those of STF. In addition, the TREAT upgrade provides for use of STF-like stainless steel-UO{sub 2} TREAT fuel for tests of fully enriched fuel bundles. This report will introduce the Upgrade study by presenting a brief description of the scope, performance capability, safety considerations, cost schedule, and development requirements. This work is followed by a "Design Description". Because greatly upgraded loop performance is central to the upgrade, a description is given of Advanced TREAT loop requirements prior to description of the loop concept. Performance requirements of the upgraded reactor system are given. An extensive discussion of the reactor physics calculations performed for the Upgrade concept study is provided. Adequate physics performance is essential for performance of experiments with the Advanced TREAT loop, and the stress placed on these calculations reflects this. Additional material on performance and safety is provided. Backup calculations on calculations of plutonium-release limits are described. Cost and schedule information for the Upgrade are presented.« 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.

Advanced Fuels Campaign FY 2015 Accomplishments Report

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

Braase, Lori Ann; Carmack, William Jonathan

2015-10-29

The mission of the Advanced Fuels Campaign (AFC) is to perform research, development, and demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This report is a compilation of technical accomplishment summaries for FY-15. Emphasis is on advanced accident-tolerant LWR fuel systems, advanced transmutation fuels technologies, and capability development.

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

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

Lisowski, Darius D.; Kraus, Adam R.; Bucknor, Matthew D.

A 1/2 scale test facility has been constructed at Argonne National Laboratory to study the heat removal performance and natural circulation flow patterns in a Reactor Cavity Cooling System (RCCS). Our test facility, the Natural convection Shutdown heat removal Test Facility (NSTF), supports the broader goal of developing an inherently safe and fully passive ex-vessel decay heat removal for advanced reactor designs. The project, initiated in 2010 to support the Advanced Reactor Technologies (ART), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs, has been conducting experimental operations since early 2014. The following paper provides a summary ofmore » some primary design features of the 26-m tall test facility along with a description of the data acquisition suite that guides our experimental practices. Specifics of the distributed fiber optic temperature measurements will be discussed, which introduces an unparalleled level of data density that has never before been implemented in a large scale natural circulation test facility. Results from our test series will then be presented, which provide insight into the thermal hydraulic behavior at steady-state and transient conditions for varying heat flux levels and exhaust chimney configuration states. (C) 2016 Elsevier B.V. All rights reserved.« less

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

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.

We report that many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has beenmore » examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Lastly, although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.« less

Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

DOE PAGES

Skavdahl, Isaac; Utgikar, Vivek; Christensen, Richard; ...

2016-05-24

We present an alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T co) and the hot outlet temperature of the intermediate heat exchanger (T ho2) by manipulating the hot-side flow rates of the heat exchangers (F h/F h2) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the controlmore » of the cold outlet temperature of the SHX (T co) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The final option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.« less

Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report

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

Jiao, Zhujie; Was, Gary; Bartels, David

2015-04-02

This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that themore » effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.« less

Principles and Application of Magnetic Rubber Testing for Crack Detection in High-Strength Steel Components: I. Active-Field Inspection

DTIC Science & Technology

2014-12-01

Historically, MRT found its most extensive application in the inspection of critical high-strength steel components of the F-111 aircraft to...Steve Burke is Group Leader Acoustic Material Systems within Maritime Division and Task Leader for AIR 07/101 Assessment and Control of Aircraft ...Maritime Division. He has previously led research programs in advanced electromagnetic and ultrasonic NDE for aircraft applications. Geoff has BSc and BE

A New Division of Labor: Meeting America’s Security Challenges Beyond Iraq

DTIC Science & Technology

2007-01-01

on North Korea’s nuclear capabilities, see Graham and Kessler (2005). 20 A New Division of Labor most populous country, China has over the past... Seth Jones, Rollie Lal, Andrew Rathmell, and Anga Timilsina, America’s Role in Nation-Building: From Germany to Iraq, Santa Monica, Calif.: RAND...Graham, Bradley, and Glenn Kessler , “N. Korea Nuclear Advance is Cited,” Washington Post, April 29, 2005, p. 1. Bibliography 107 Headquarters

NASA Planetary Science Division's Instrument Development Programs, PICASSO and MatISSE

NASA Technical Reports Server (NTRS)

Gaier, James R.

2016-01-01

The Planetary Science Division (PSD) has combined several legacy instrument development programs into just two. The Planetary Instrument Concepts Advancing Solar System Observations (PICASSO) program funds the development of low TRL instruments and components. The Maturation of Instruments for Solar System Observations (MatISSE) program funds the development of instruments in the mid-TRL range. The strategy of PSD instrument development is to develop instruments from PICASSO to MatISSE to proposing for mission development.

Determination of the Impregnant Concentrations on ASC Type Charcoal. A Magnetic Susceptibility Study

DTIC Science & Technology

1987-10-01

commercial magnetic susceptometers were used on site at the Geological Division of CANMET , whose help is greatly appreciated. One of the susceptometers was a...CHARCOAL. A MAGNETIC SUSCEPTIBILITY STUDY (U) by S.H.C. Liang, B.H. Harrison and J.G. Pagotto Chemical Protection Section Protective Sciences Division ...analysis (11, 16-18), and surveys of literature reports (19-23). Some recent advances in instrumentation, such as the vibrating sample magnetometer (24

General Troy H. Middleton: Steadfast in Command

DTIC Science & Technology

2011-05-19

front, flanks and rear open…Bradley simply wants a bridgehead over the Selune River. What I want and intend to get is Brest and Angers.”77 Patton...during the Normandy breakout. He had ordered the 6th Armored Division to advance on Brest . However, Patton found Grow’s 6th Armored Division...approximately a hundred miles from Brest on August 3, 1944.78 Patton reviewed Middleton’s orders that assigned an intermediate objective of Dinan en route to St

Sensitivity Analysis and Optimization of the Nuclear Fuel Cycle: A Systematic Approach

NASA Astrophysics Data System (ADS)

Passerini, Stefano

For decades, nuclear energy development was based on the expectation that recycling of the fissionable materials in the used fuel from today's light water reactors into advanced (fast) reactors would be implemented as soon as technically feasible in order to extend the nuclear fuel resources. More recently, arguments have been made for deployment of fast reactors in order to reduce the amount of higher actinides, hence the longevity of radioactivity, in the materials destined to a geologic repository. The cost of the fast reactors, together with concerns about the proliferation of the technology of extraction of plutonium from used LWR fuel as well as the large investments in construction of reprocessing facilities have been the basis for arguments to defer the introduction of recycling technologies in many countries including the US. In this thesis, the impacts of alternative reactor technologies on the fuel cycle are assessed. Additionally, metrics to characterize the fuel cycles and systematic approaches to using them to optimize the fuel cycle are presented. The fuel cycle options of the 2010 MIT fuel cycle study are re-examined in light of the expected slower rate of growth in nuclear energy today, using the CAFCA (Code for Advanced Fuel Cycle Analysis). The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycle options available in the future. The options include limited recycling in L WRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. Additional fuel cycle scenarios presented for the first time in this work assume the deployment of innovative recycling reactor technologies such as the Reduced Moderation Boiling Water Reactors and Uranium-235 initiated Fast Reactors. A sensitivity study focused on system and technology parameters of interest has been conducted to test the robustness of the conclusions presented in the MIT Fuel Cycle Study. These conclusions are found to still hold, even when considering alternative technologies and different sets of simulation assumptions. Additionally, a first of a kind optimization scheme for the nuclear fuel cycle analysis is proposed and the applications of such an optimization are discussed. Optimization metrics of interest for different stakeholders in the fuel cycle (economics, fuel resource utilization, high level waste, transuranics/proliferation management, and environmental impact) are utilized for two different optimization techniques: a linear one and a stochastic one. Stakeholder elicitation provided sets of relative weights for the identified metrics appropriate to each stakeholder group, which were then successfully used to arrive at optimum fuel cycle configurations for recycling technologies. The stochastic optimization tool, based on a genetic algorithm, was used to identify non-inferior solutions according to Pareto's dominance approach to optimization. The main tradeoff for fuel cycle optimization was found to be between economics and most of the other identified metrics. (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

AICD -- Advanced Industrial Concepts Division Biological and Chemical Technologies Research Program. 1993 Annual summary report

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

Petersen, G.; Bair, K.; Ross, J.

1994-03-01

The annual summary report presents the fiscal year (FY) 1993 research activities and accomplishments for the United States Department of Energy (DOE) Biological and Chemical Technologies Research (BCTR) Program of the Advanced Industrial Concepts Division (AICD). This AICD program resides within the Office of Industrial Technologies (OIT) of the Office of Energy Efficiency and Renewable Energy (EE). The annual summary report for 1993 (ASR 93) contains the following: A program description (including BCTR program mission statement, historical background, relevance, goals and objectives), program structure and organization, selected technical and programmatic highlights for 1993, detailed descriptions of individual projects, a listingmore » of program output, including a bibliography of published work, patents, and awards arising from work supported by BCTR.« less

Control console replacement at the WPI Reactor. [Final report

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

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

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

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

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

2016-08-31

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

VERA Core Simulator methodology for pressurized water reactor cycle depletion

DOE PAGES

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

2017-01-12

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

NACA Researcher Examines the Cyclotron

NASA Image and Video Library

1951-02-21

Researcher James Blue examines the new cyclotron at the National Advisory Committee for Aeronautics (NACA) Lewis Flight Propulsion Laboratory. Researchers at NACA Lewis began postulating about the use of atomic power for propulsion immediately after World War II. The NACA concentrated its efforts on the study of high temperature materials and heat transfer since it did not have access to the top secret fission information. The military studied the plausibility of nuclear propulsion for aircraft in the late 1940s. The military program was cancelled after four years without any breakthroughs, but the Atomic Energy Commission took on the effort in 1951. The NACA Lewis laboratory was expanding its nuclear-related research during this period. In 1948, Lewis engineers were assigned to the Oak Ridge National Laboratory to obtain expertise in high temperature heat transfer and advanced materials technology. The following year a new 80-person Nuclear Reactor Division was created, and an in-house nuclear school was established to train these researchers. The cyclotron was built behind the Materials and Structures Laboratory to support thermodynamic and materials research for both nuclear aircraft and nuclear rockets. The original NACA Lewis cyclotron was used to accelerate two kinds of particles. To better match the space radiation environment, the cyclotron was later modified to accelerate particles of the newly-discovered Van Allen radiation belts.

A Survey of Past Work on Rates of Advance in Land Combat Operations

DTIC Science & Technology

1990-02-01

Weather: Clear, mild, cold, snow, fog, rain, misty , fair-again objective definitions are not provided. c. Type action: Pursuit, limited objective...90-03 SIMPKIN-1984 J 1. Document: a. Title: Red Armour b. Author: Simpkin, Richard E. c. Date: 1984 d. Organiation: Pergamon Press Ltd, Oxford e...8217Mechanised advances on roads average about 35 miles,’day, advances on the best days achieving 50 miles for typical armoured division wheel/track mixes

Advanced planetary studies

NASA Technical Reports Server (NTRS)

1982-01-01

Results of planetary advanced studies and planning support provided by Science Applications, Inc. staff members to Earth and Planetary Exploration Division, OSSA/NASA, for the period 1 February 1981 to 30 April 1982 are summarized. The scope of analyses includes cost estimation, planetary missions performance, solar system exploration committee support, Mars program planning, Galilean satellite mission concepts, and advanced propulsion data base. The work covers 80 man-months of research. Study reports and related publications are included in a bibliography section.

Publications - Advanced Search | Alaska Division of Geological &

Science.gov Websites

publications released by the University of Alaska Fairbanks, Mineral Industry Research Laboratory USBM Island Seguam Selawik Seldovia Seward Shishmaref Shungnak Simeonof Island Sitka Skagway Sleetmute Solomon

Advanced Code-Division Multiplexers for Superconducting Detector Arrays

NASA Astrophysics Data System (ADS)

Irwin, K. D.; Cho, H. M.; Doriese, W. B.; Fowler, J. W.; Hilton, G. C.; Niemack, M. D.; Reintsema, C. D.; Schmidt, D. R.; Ullom, J. N.; Vale, L. R.

2012-06-01

Multiplexers based on the modulation of superconducting quantum interference devices are now regularly used in multi-kilopixel arrays of superconducting detectors for astrophysics, cosmology, and materials analysis. Over the next decade, much larger arrays will be needed. These larger arrays require new modulation techniques and compact multiplexer elements that fit within each pixel. We present a new in-focal-plane code-division multiplexer that provides multiplexing elements with the required scalability. This code-division multiplexer uses compact lithographic modulation elements that simultaneously multiplex both signal outputs and superconducting transition-edge sensor (TES) detector bias voltages. It eliminates the shunt resistor used to voltage bias TES detectors, greatly reduces power dissipation, allows different dc bias voltages for each TES, and makes all elements sufficiently compact to fit inside the detector pixel area. These in-focal plane code-division multiplexers can be combined with multi-GHz readout based on superconducting microresonators to scale to even larger arrays.

ER-mitochondria contacts couple mtDNA synthesis with mitochondrial division in human cells.

PubMed

Lewis, Samantha C; Uchiyama, Lauren F; Nunnari, Jodi

2016-07-15

Mitochondrial DNA (mtDNA) encodes RNAs and proteins critical for cell function. In human cells, hundreds to thousands of mtDNA copies are replicated asynchronously, packaged into protein-DNA nucleoids, and distributed within a dynamic mitochondrial network. The mechanisms that govern how nucleoids are chosen for replication and distribution are not understood. Mitochondrial distribution depends on division, which occurs at endoplasmic reticulum (ER)-mitochondria contact sites. These sites were spatially linked to a subset of nucleoids selectively marked by mtDNA polymerase and engaged in mtDNA synthesis--events that occurred upstream of mitochondrial constriction and division machine assembly. Our data suggest that ER tubules proximal to nucleoids are necessary but not sufficient for mtDNA synthesis. Thus, ER-mitochondria contacts coordinate licensing of mtDNA synthesis with division to distribute newly replicated nucleoids to daughter mitochondria. Copyright © 2016, American Association for the Advancement of Science.

Milestone M3FT-15OR0203112. Build redesigned HFIR rabbit capsules and make ready for insertion for irradiation in HFIR

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

Howard, Richard H; McDuffee, Joel Lee; Okuniewski, Maria A.

2015-09-01

This report details the fabrication and delivery of two Fuel Cycle Research and Development irradiation capsules (FCRP20 and FCRP03), with associated quality assurance documentation, to the High Flux Isotope Reactor. The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0203112. These irradiation experiments irradiate metal parallelepiped specimens that may consist of various compositions including uranium metal, steel, etc. This document contains a copy of the completed capsule fabrication request sheets, which detail all constituent components, pertinent drawings, etc., along with a detailed summary of the capsule assembly process performed by the Thermal Hydraulicsmore » and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division. A complete fabrication package record is maintained by THIEG and is available upon request.« less

The ENABLER—based on proven NERVA technology

NASA Astrophysics Data System (ADS)

Livingston, Julie M.; Pierce, Bill L.

1991-01-01

The ENABLER reactor for use in a nuclear thermal propulsion engine uses the technology developed in the NERVA/Rover program, updated to incorporate advances in the technology. Using composite fuel, higher power densities per fuel element, improved radiation resistant control components and the advancements in use of carbon-carbon materials; the ENABLER can provide a specific impulse of 925 seconds, an engine thrust to weight (excluding reactor shield) approaching five, an improved initial Mass In Low Earth Orbit (IMLEO) and a consequent reduction in launch costs and logistics problems. This paper describes the 75,000 lbs thrust ENABLER design which is a low cost, low risk approach to meeting tomorrow's space propulsion needs.

Initial Back-to-Back Fission Chamber Testing in ATRC

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

Benjamin Chase; Troy Unruh; Joy Rempe

2014-06-01

Development and testing of in-pile, real-time neutron sensors for use in Materials Test Reactor experiments is an ongoing project at Idaho National Laboratory. The Advanced Test Reactor National Scientific User Facility has sponsored a series of projects to evaluate neutron detector options in the Advanced Test Reactor Critical Facility (ATRC). Special hardware was designed and fabricated to enable testing of the detectors in the ATRC. Initial testing of Self-Powered Neutron Detectors and miniature fission chambers produced promising results. Follow-on testing required more experiment hardware to be developed. The follow-on testing used a Back-to-Back fission chamber with the intent to providemore » calibration data, and a means of measuring spectral indices. As indicated within this document, this is the first time in decades that BTB fission chambers have been used in INL facilities. Results from these fission chamber measurements provide a baseline reference for future measurements with Back-to-Back fission chambers.« less

Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

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

J. Benson; J. Cole; J. Jackson

2013-02-01

In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groupsmore » conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.« less

Feasibility of EB Welded Hastelloy X and Combination of Refractory Metals

NASA Technical Reports Server (NTRS)

Martinez, Diana A.

2004-01-01

As NASA continues to expand its horizon, exploration and discovery creates the need of advancement in technology. The Jupiter Icy Moon Orbiter's (JIMO) mission to explore and document the outer surfaces, rate the possibility of holding potential life forms, etc. within the three moons (Callisto, Ganymede, and Europa) proves to be challenging. The orbiter itself consists of many sections including: the nuclear reactor and the power conversion system, the radiator panels, and the thrusters and antenna. The nuclear reactor serves as a power source, and if successfully developed, can operate for extended periods. During the duration of my tenure at NASA Glenn Research Center's (NASA GRC) Advanced Metallics Branch, I was assigned to assist Frank J. Ritzert on analyzing the feasibility of the Electron Beam Welded Hastelloy X (HX), a nickel-based superalloy, to Niobium- 1 %Zirconium (Nb-1 Zr) and other refractory metals/alloys including Tantalum, Molybdenum, Tungsten, and Rhenium alloys. This welding technique is going to be used for the nuclear reactor within JIMO.

TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

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

Not Available

1978-05-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions.more » The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.« less

AGR-2 and AGR-3/4 Release-to-Birth Ratio Data Analysis

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

Pham, Binh T.; Einerson, Jeffrey J.; Scates, Dawn M.

A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low enriched fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples embedded in the graphite enabling temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through use of heliummore » coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO coated particle fuel. Thus, these tests provide valuable irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, and support development and validation of fuel performance and fission product transport models and codes.« less

High Efficiency Nuclear Power Plants using Liquid Fluoride Thorium Reactor Technology

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

2009-01-01

An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITS of 950 K and 1200 K are presented. Power plant performance data were obtained for TITS ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo -generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit

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

Merzari, E.; Shemon, E. R.; Yu, Y. Q.

This report describes to employ SHARP to perform a first-of-a-kind analysis of the core radial expansion phenomenon in an SFR. This effort required significant advances in the framework Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit used to drive the coupled simulations, manipulate the mesh in response to the deformation of the geometry, and generate the necessary modified mesh files. Furthermore, the model geometry is fairly complex, and consistent mesh generation for the three physics modules required significant effort. Fully-integrated simulations of a 7-assembly mini-core test problem have been performed, and the results are presented here. Physics models ofmore » a full-core model of the Advanced Burner Test Reactor have also been developed for each of the three physics modules. Standalone results of each of the three physics modules for the ABTR are presented here, which provides a demonstration of the feasibility of the fully-integrated simulation.« less