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

Polymerization Reactor Engineering.

ERIC Educational Resources Information Center

Skaates, J. Michael

1987-01-01

Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

75 FR 4493 - Natural Resources Defense Council; Denial of Petition for Rulemaking

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-28

... NRC continues to license the civilian use of HEU to fuel seven existing research and test reactors... predicts that the three HEU-fueled TRIGA-type research reactors at Oregon State University, the University...) is scheduled for conversion to LEU but notes that the newer and larger LEU-fueled TRIGA facility at...

Plasma Physics Lab and the Tokamak Fusion Test Reactor, 1989

ScienceCinema

None

2018-01-16

From the Princeton University Archives: Promotional video about the Plasma Physics Lab and the new Tokamak Fusion Test Reactor (TFTR), with footage of the interior, machines, and scientists at work. This film is discussed in the audiovisual blog of the Seeley G. Mudd Manuscript Library, which holds the archives of Princeton University.

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

Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

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

Not Available

1995-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. 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 undergraduate 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 low power output of the reactor and an ergonomicmore » facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.« less

75 FR 54657 - University of Florida; University of Florida Training Reactor; Environmental Assessment and...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-08

... operation of the UFTR to routinely provide teaching, research, and services to numerous institutions for a... confinement. The Nuclear Reactor Building and its annex, the Nuclear Sciences Center, are located in an area... primary system consisting of a 200-gallon coolant storage tank, a heat removal system, and a processing...

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

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

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

Chemical Reactors.

ERIC Educational Resources Information Center

Kenney, C. N.

1980-01-01

Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

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

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

2009-05-01

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

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

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

The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Programmore » since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993.« less

An Integrated Fuel Depletion Calculator for Fuel Cycle Options Analysis

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

Schneider, Erich; Scopatz, Anthony

2016-04-25

Bright-lite is a reactor modeling software developed at the University of Texas Austin to expand upon the work done with the Bright [1] reactor modeling software. Originally, bright-lite was designed to function as a standalone reactor modeling software. However, this aim was refocused t couple bright-lite with the Cyclus fuel cycle simulator [2] to make it a module for the fuel cycle simulator.

A Potential NASA Research Reactor to Support NTR Development

NASA Technical Reports Server (NTRS)

Eades, Michael; Gerrish, Harold; Hardin, Leroy

2013-01-01

In support of efforts for research into the design and development of a man rated Nuclear Thermal Rocket (NTR) engine, the National Aeronautics and Space Administration (NASA), Marshall Space Flight Center (MSFC), is evaluating the potential for building a Nuclear Regulatory Commission (NRC) licensed research reactor. The proposed reactor would be licensed by NASA and operated jointly by NASA and university partners. The purpose of this reactor would be to perform further research into the technologies and systems needed for a successful NTR project and promote nuclear training and education.

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.

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

NASA Astrophysics Data System (ADS)

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

2017-08-01

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

An on-line reactivity and power monitor for a TRIGA reactor

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

Binney, Stephen E.; Bakir, Alia J.

1988-07-01

As the personal computer (PC) becomes more and more of a significant influence on modern technology, it is reasonable that at some point in time they would be used to interface with TRIGA reactors. A personal computer with a special interface board has been used to monitor key parameters during operation of the Oregon State University TRIGA Reactor (OSTR). A description of the apparatus used and sample results are included.

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

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

Ryan, B.C.

1997-05-01

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

Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration

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

J. Wesley Hines; Belle R. Upadhyaya; J. Michael Doster

2011-05-31

Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topicalmore » areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor deployment analysis and small grid stability issues. Objective 2 - Develop and test an autonomous and fault-tolerant control architecture and apply to the IRIS system and an experimental flow control loop, with extensions to multiple reactor modules, nuclear desalination, and optimal sensor placement strategy. Objective 3 - Develop and test an integrated monitoring, diagnosis, and prognosis system for SMRs using the IRIS as a test platform, and integrate process and equipment monitoring (PEM) and process and equipment prognostics (PEP) toolboxes. The research tasks are focused on meeting the unique needs of reactors that may be deployed to remote locations or to developing countries with limited support infrastructure. These applications will require smaller, robust reactor designs with advanced technologies for sensors, instrumentation, and control. An excellent overview of SMRs is described in an article by Ingersoll (2009). The article refers to these as deliberately small reactors. Most of these have modular characteristics, with multiple units deployed at the same plant site. Additionally, the topics focus on meeting two of the eight needs outlined in the recently published 'Technology Roadmap on Instrumentation, Control, and Human-Machine Interface (ICHMI) to Support DOE Advanced Nuclear Energy Programs' which was created 'to provide a systematic path forward for the integration of new ICHMI technologies in both near-term and future nuclear power plants and the reinvigoration of the U.S. nuclear ICHMI community and capabilities.' The research consortium is led by The University of Tennessee (UT) and is focused on three interrelated topics: Topic 1 (simulator development and measurement sensitivity analysis) is led by Dr. Mike Doster with Dr. Paul Turinsky of North Carolina State University (NCSU). Topic 2 (multivariate autonomous control of modular reactors) is led by Dr. Belle Upadhyaya of the University of Tennessee (UT) and Dr. Robert Edwards of Penn State University (PSU). Topic 3 (monitoring, diagnostics, and prognostics system development) is led by Dr. Wes Hines of UT. Additionally, South Carolina State University (SCSU, Dr. Ken Lewis) participated in this research through summer interns, visiting faculty, and on-campus research projects identified throughout the grant period. Lastly, Westinghouse Science and Technology Center (Dr. Mario Carelli) was a no-cost collaborator and provided design information related to the IRIS demonstration platform and defining needs that may be common to other SMR designs. The results of this research are reported in a six-volume Final Report (including the Executive Summary, Volume 1). Volumes 2 through 6 of the report describe in detail the research and development under the topical areas. This volume serves to introduce the overall NERI-C project and to summarize the key results. Section 2 provides a summary of the significant contributions of this project. A list of all the publications under this project is also given in Section 2. Section 3 provides a brief summary of each of the five volumes (2-6) of the report. The contributions of SCSU are described in Section 4, including a summary of undergraduate research experience. The project management organizational chart is provided as Figure 1. Appendices A, B, and C contain the reports on the summer research performed at the University of Tennessee by undergraduate students from South Carolina State University.« 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

Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

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

Scervini, M.; Palmer, J.; Haggard, D.C.

2015-07-01

Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation formore » relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of Cambridge have been investigated. The rationale for the superior performance of the type N using a customized sheath developed at the University of Cambridge is explained in comparison with the behavior of conventional type N Inconel 600 sheathed thermocouples. (authors)« less

Modifications to the NRAD Reactor, 1977 to present

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

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

1986-01-01

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

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

Edgue, E.

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

Fundamental interactions involving neutrons and neutrinos: reactor-based studies led by Petersburg Nuclear Physics Institute (National Research Centre 'Kurchatov Institute') [PNPI (NRC KI)

NASA Astrophysics Data System (ADS)

Serebrov, A. P.

2015-11-01

Neutrons of very low energy ( ˜ 10-7 eV), commonly known as ultracold, are unique in that they can be stored in material and magnetic traps, thus enhancing methodical opportunities to conduct precision experiments and to probe the fundamentals of physics. One of the central problems of physics, of direct relevance to the formation of the Universe, is the violation of time invariance. Experiments searching for the nonzero neutron electric dipole moment serve as a time invariance test, and the use of ultracold neutrons provides very high measurement precision. Precision neutron lifetime measurements using ultracold neutrons are extremely important for checking ideas on the early formation of the Universe. This paper discusses problems that arise in studies using ultracold neutrons. Also discussed are the currently highly topical problem of sterile neutrinos and the search for reactor antineutrino oscillations at distances of 6-12 meters from the reactor core. The field reviewed is being investigated at multiple facilities globally. The present paper mainly concentrates on the results of PNPI-led studies at WWR-M PNPI (Gatchina), ILL (Grenoble), and SM-3 (Dimitrovgrad) reactors, and also covers the results obtained during preparation for research at the PIK reactor which is under construction.

75 FR 27372 - University of New Mexico; University of New Mexico AGN-201M Reactor; Environmental Assessment and...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-14

... is housed in the Nuclear Energy Laboratory (NEL) located near the southwest corner of the University.... Therefore, license renewal should not change the environmental impact of facility operation. Data from the... analysis of human health and environmental impacts presented in this environmental assessment, the proposed...

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

Total absorption studies of high priority decays for reactor applications: 86 Br and 91 Rb

DOE PAGES

Algora, A.; Rice, S.; Guadilla, V.; ...

2017-09-13

Preliminary results from beta decay studies of nuclei that are important for reactor applications are presented. The beta decays have been studied using the total absorption technique (TAS) and the pure beams provided by the JYFLTRAP system at the IGISOL facility of the University of Jyväskylä.

Graphite for the nuclear industry

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

Burchell, T.D.; Fuller, E.L.; Romanoski, G.R.

Graphite finds applications in both fission and fusion reactors. Fission reactors harness the energy liberated when heavy elements, such as uranium or plutonium, fragment or fission''. Reactors of this type have existed for nearly 50 years. The first nuclear fission reactor, Chicago Pile No. 1, was constructed of graphite under a football stand at Stagg Field, University of Chicago. Fusion energy devices will produce power by utilizing the energy produced when isotopes of the element hydrogen are fused together to form helium, the same reaction that powers our sun. The role of graphite is very different in these two reactormore » systems. Here we summarize the function of the graphite in fission and fusion reactors, detailing the reasons for their selection and discussing some of the challenges associated with their application in nuclear fission and fusion reactors. 10 refs., 15 figs., 1 tab.« less

Expanded scope of training and education programs at the UFTR

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

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

1985-01-01

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

Research reactor decommissioning experience - concrete removal and disposal -

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

Manning, Mark R.; Gardner, Frederick W.

1990-07-01

Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limitsmore » for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations.« less

78 FR 29393 - University of Missouri-Columbia Facility Operating License No. R-103

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-20

... Curators of the University of Missouri--Columbia (the licensee) to operate the Missouri University Research Reactor (MURR) at a maximum steady-state thermal power of 10 megawatts (MW). The renewed license would authorize the licensee to operate the MURR up to a steady-state thermal power of 10 MW for an additional 20...

Modeling of biomass to hydrogen via the supercritical water pyrolysis process

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

Divilio, R.J.

1998-08-01

A heat transfer model has been developed to predict the temperature profile inside the University of Hawaii`s Supercritical Water Reactor. A series of heat transfer tests were conducted on the University of Hawaii`s apparatus to calibrate the model. Results of the model simulations are shown for several of the heat transfer tests. Tests with corn starch and wood pastes indicated that there are substantial differences between the thermal properties of the paste compared to pure water, particularly near the pseudo critical temperature. The assumption of constant thermal diffusivity in the temperature range of 250 to 450 C gave a reasonablemore » prediction of the reactor temperatures when paste is being fed. A literature review is presented for pyrolysis of biomass in water at elevated temperatures up to the supercritical range. Based on this review, a global reaction mechanism is proposed. Equilibrium calculations were performed on the test results from the University of Hawaii`s Supercritical Water Reactor when corn starch and corn starch and wood pastes were being fed. The calculations indicate that the data from the reactor falls both below and above the equilibrium hydrogen concentrations depending on test conditions. The data also indicates that faster heating rates may be beneficial to the hydrogen yield. Equilibrium calculations were also performed to examine the impact of wood concentration on the gas mixtures produced. This calculation showed that increasing wood concentrations favors the formation of methane at the expense of hydrogen.« less

Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators

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

Palabrica, R.J.

1981-01-01

The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

Operations of a TRIGA reactor at a small private liberal arts college

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

Church, L.B.

A small private liberal arts college is not a very representative place to have a TRIGA reactor. Reed is a wholly undergraduate institution with a strong emphasis in the traditional liberal arts and fundamental sciences. Many of the larger state universities provide an excellence in nuclear science which is often presented to students in a somewhat distant manner. By providing a reactor that was immediately accessible to undergraduate students it has been realized that the excitement attendant with nuclear science would be available to them in an immediate hands-on manner.

Plasma reactor waste management systems

NASA Technical Reports Server (NTRS)

Ness, Robert O., Jr.; Rindt, John R.; Ness, Sumitra R.

1992-01-01

The University of North Dakota is developing a plasma reactor system for use in closed-loop processing that includes biological, materials, manufacturing, and waste processing. Direct-current, high-frequency, or microwave discharges will be used to produce plasmas for the treatment of materials. The plasma reactors offer several advantages over other systems, including low operating temperatures, low operating pressures, mechanical simplicity, and relatively safe operation. Human fecal material, sunflowers, oats, soybeans, and plastic were oxidized in a batch plasma reactor. Over 98 percent of the organic material was converted to gaseous products. The solids were then analyzed and a large amount of water and acid-soluble materials were detected. These materials could possibly be used as nutrients for biological systems.

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

Stillman, J. A.; Feldman, E. E.; Jaluvka, D.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclearmore » Security Administration’s Office of Material Management and Minimization (M 3).« less

The WPI reactor-readying for the next generation

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

Bobek, L.M.

1993-01-01

Built in 1959, the 10-kW open-pool nuclear training reactor at Worcester Polytechnic Institute (WPI) was one of the first such facilities in the nation located on a university campus. Since then, the reactor and its related facilities have been used to train two generations of nuclear engineers and scientists for the nuclear industry. With the use of nuclear technology playing an increasing role in many segments of the economy, WPI with its nuclear reactor facility is committed to continuing its mission of training future nuclear engineers and scientists. The WPI reactor includes a 6-in. beam port, graphite thermal column, andmore » in-core sample facility. The reactor, housed in an open 8000-gal tank of water, is designed so that the core is readily accessible. Both the control console and the peripheral counting equipment used for student projects and laboratory exercises are located in the reactor room. This arrangement provides convenience and flexibility in using the reactor for foil activations in neutron flux measurements, diffusion measurements, radioactive decay measurements, and the neutron activation of samples for analysis. In 1988, the reactor was successfully converted to low-enriched uranium fuel.« less

Element 117

ScienceCinema

None

2018-06-12

An international team of scientists from Russia and the United States, including two Department of Energy national laboratories and two universities, has discovered the newest superheavy element, element 117. The team included scientists from the Joint Institute of Nuclear Research (Dubna, Russia), the Research Institute for Advanced Reactors (Dimitrovgrad), Lawrence Livermore National Laboratory, Oak Ridge National Laboratory, Vanderbilt University, and the University of Nevada, Las Vegas.

Background studies for the MINER Coherent Neutrino Scattering reactor experiment

NASA Astrophysics Data System (ADS)

Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T. J.; Cesar, J.; Cushman, P.; Dent, J. B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H. R.; Hays, C. C.; Iyer, V.; Jastram, A.; Kadribasic, F.; Kennedy, A.; Kubik, A.; Lang, K.; Mahapatra, R.; Mandic, V.; Marianno, C.; Martin, R. D.; Mast, N.; McDeavitt, S.; Mirabolfathi, N.; Mohanty, B.; Nakajima, K.; Newhouse, J.; Newstead, J. L.; Ogawa, I.; Phan, D.; Proga, M.; Rajput, A.; Roberts, A.; Rogachev, G.; Salazar, R.; Sander, J.; Senapati, K.; Shimada, M.; Soubasis, B.; Strigari, L.; Tamagawa, Y.; Teizer, W.; Vermaak, J. I. C.; Villano, A. N.; Walker, J.; Webb, B.; Wetzel, Z.; Yadavalli, S. A.

2017-05-01

The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5-20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

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

Hetrick, D.L.; Sowers, G.W.

1978-06-01

This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. Amore » list of variable names and a listing for BRENDA are included as appendices.« less

Microbial community composition is consistent across anaerobic digesters processing wheat-based fuel ethanol waste streams.

PubMed

Town, Jennifer; Annand, Holly; Pratt, Dyan; Dumonceaux, Tim; Fonstad, Terrance

2014-04-01

Biochemical methane potential (BMP) assays were conducted on byproducts from dry-grind wheat-based ethanol plants amended with feedlot manure at two input ratios. Whole stillage (WST), thin stillage (TST) and wet cake (WCK) were tested alone and with 1:1 and 2:1 ratios (VS basis) of byproduct:feedlot manure in bench-scale batch reactors. The addition of manure increased both the rate and consistency of methane production in triplicate reactors. In addition, digesters co-digesting thin stillage and cattle manure at 1:1 and 2:1 stillage:manure produced 125% and 119% expected methane based on the biomethane potential of each substrate digested individually. Bacterial community analysis using universal target amplification and pyrosequencing indicated there was a numerically dominant core of 42 bacteria that was universally present in the reactors regardless of input material. A smaller-scale analysis of the archaeal community showed that both hydrogenotrophic and acetoclastic methanogens were present in significant quantities. Crown Copyright © 2014. Published by Elsevier Ltd. All rights reserved.

Isotope shortage triggers delays for patients

NASA Astrophysics Data System (ADS)

Gould, Paula

2009-07-01

An unplanned shutdown of a nuclear reactor in Canada is disrupting the supply of medical isotopes across North America and forcing some hospitals to cancel or postpone patients' tests. The closure of the National Research Universal (NRU) reactor in Chalk River, Ontario, has also embarrassed Canadian officials, including a senior government minister who was forced to apologize after calling the isotope shortage a "sexy" career challenge.

Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

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

Ross, Kyle W.; Gauntt, Randall O.; Cardoni, Jeffrey N.

2013-11-01

Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.

Characterization of fast neutron spectrum in the TRIGA for hardness testing of electronic components

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

Nelson, George W.

1986-07-01

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

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

NASA Astrophysics Data System (ADS)

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

2017-11-01

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

75 FR 56597 - University of Wisconsin; University of Wisconsin Nuclear Reactor Environmental Assessment and...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-16

...,000 gallons being typical. The licensee maintains a pool leak surveillance program. The pool water leak surveillance program continues to monitor the pool water evaporation rate, the pool water make-up volume, and pool water radioactivity. The pool leak surveillance program indicated that approximately 2...

Generic Stellarator-like Magnetic Fusion Reactor

NASA Astrophysics Data System (ADS)

Sheffield, John; Spong, Donald

2015-11-01

The Generic Magnetic Fusion Reactor paper, published in 1985, has been updated, reflecting the improved science and technology base in the magnetic fusion program. Key changes beyond inflation are driven by important benchmark numbers for technologies and costs from ITER construction, and the use of a more conservative neutron wall flux and fluence in modern fusion reactor designs. In this paper the generic approach is applied to a catalyzed D-D stellarator-like reactor. It is shown that an interesting power plant might be possible if the following parameters could be achieved for a reference reactor: R/ < a > ~ 4 , confinement factor, fren = 0.9-1.15, < β > ~ 8 . 0 -11.5 %, Zeff ~ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ~ 0.07, Bm ~ 14-16 T, and R ~ 18-24 m. J. Sheffield was supported under ORNL subcontract 4000088999 with the University of Tennessee.

Research Program of a Super Fast Reactor

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

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

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

Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

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

Anderson, Mark; Sridharan, Kumar; Morgan, Dane

2015-01-22

The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsinmore » had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re-evaluate thermophysical properties of flibe and flinak. Pacific Northwest National Laboratories has focused on evaluating the fluorinating gas nitrogen trifluoride as a potential salt purification agent. Work there was performed on removing hydroxides and oxides from flinak salt under controlled conditions. Lastly, the University of California Berkeley has spent considerable time designing and simulating reactor components with fluoride salts at high temperatures. Despite the hurdles presented by the innate chemical hazards, considerable progress has been made. The stage has been set to perform new research on salt chemical control which could advance the fluoride salt cooled reactor concept towards commercialization. What were previously thought of as chemical undesirable, but nuclear certified, alloys have been shown to be theoretically compatible with fluoride salts at high temperatures. This preliminary report has been prepared to communicate the construction of the basic infrastructure required for flibe, as well as suggest original research to performed at the University of Wisconsin. Simultaneously, the contents of this report can serve as a detailed, but introductory guide to allow anyone to learn the fundamentals of chemistry, engineering, and safety required to work with flibe salt.« less

Safety analysis

NASA Technical Reports Server (NTRS)

Knight, John C.

1995-01-01

We are engaged in a research program in safety-critical computing that is based on two case studies. We use these case studies to provide application-specific details of the various research issues, and as targets for evaluation of research ideas. The first case study is the Magnetic Stereotaxis System (MSS), an investigational device for performing human neurosurgery being developed in a joint effort between the Department of Physics at the University of Virginia and the Department of Neurosurgery at the University of Iowa. The system operates by manipulating a small permanent magnet (known as a 'seed') within the brain using an externally applied magnetic field. By varying the magnitude and gradient of the external magnetic field, the seed can be moved along a non-linear path and positioned at a site requiring therapy, e.g., a tumor. The magnetic field required for movement through brain tissue is extremely high, and is generated by a set of six superconducting magnets located in a housing surrounding the patient's head. The system uses two X-ray cameras positioned at right angles to detect in real time the locations of the seed and of X-ray opaque markers affixed to the patient's skull. the X-ray images are used to locate the objects of interest in a canonical frame of reference. the second case study is the University of Virginia Research Nuclear Reactor (UVAR). It is a 2 MW thermal, concrete-walled pool reactor. The system operates using 20 to 25 plate-type fuel assemblies placed on a rectangular grid plate. There are three scramable safety rods, and one non-scramable regulating rod that can be put in automatic mode. It was originally constructed in 1959 as a 1 MW system, and it was upgraded to 2 MW in 1973. Though only a research reactor rather than a power reactor, the issues raised are significant and can be related to the problems faced by full-scale reactor systems.

Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

NASA Astrophysics Data System (ADS)

Hu, Jian; Jiang, Nan; Li, Jie; Shang, Kefeng; Lu, Na; Wu, Yan; Mizuno, Akira

2016-03-01

The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. supported by National Natural Science Foundation of China (No. 51177007), the Joint Funds of National Natural Science Foundation of China (No. U1462105), and Dalian University of Technology Fundamental Research Fund of China (No. DUT15RC(3)030)

The First Reactor [40th Anniversary Commemorative Edition].

DOE R&D Accomplishments Database

1982-12-01

This updated and revised story of the first reactor, or 'pile,' commemorates the 40th anniversary of the first controlled, self-sustaining nuclear chain reaction created by mankind. Enrico Fermi and his team of scientists initiated the reaction on December 2, 1941, underneath the West Stands of Stagg Field at the University of Chicago. Firsthand accounts of the participants as well as postwar recollections by Enrico and Laura Fermi are included.

Thermal barrier coatings on gas turbine blades: Chemical vapor deposition (Review)

NASA Astrophysics Data System (ADS)

Igumenov, I. K.; Aksenov, A. N.

2017-12-01

Schemes are presented for experimental setups (reactors) developed at leading scientific centers connected with the development of technologies for the deposition of coatings using the CVD method: at the Technical University of Braunschweig (Germany), the French Aerospace Research Center, the Materials Research Institute (Tohoku University, Japan) and the National Laboratory Oak Ridge (USA). Conditions and modes for obtaining the coatings with high operational parameters are considered. It is established that the formed thermal barrier coatings do not fundamentally differ in their properties (columnar microstructure, thermocyclic resistance, thermal conductivity coefficient) from standard electron-beam condensates, but the highest growth rates and the perfection of the crystal structure are achieved in the case of plasma-chemical processes and in reactors with additional laser or induction heating of a workpiece. It is shown that CVD reactors can serve as a basis for the development of rational and more advanced technologies for coating gas turbine blades that are not inferior to standard electron-beam plants in terms of the quality of produced coatings and have a much simpler and cheaper structure. The possibility of developing a new technology based on CVD processes for the formation of thermal barrier coatings with high operational parameters is discussed, including a set of requirements for industrial reactors, high-performance sources of vapor precursors, and promising new materials.

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

Dehart, Mark; Mausolff, Zander; Goluoglu, Sedat

This report summarizes university research activities performed in support of TREAT modeling and simulation research. It is a compilation of annual research reports from four universities: University of Florida, Texas A&M University, Massachusetts Institute of Technology and Oregon State University. The general research topics are, respectively, (1) 3-D time-dependent transport with TDKENO/KENO-VI, (2) implementation of the Improved Quasi-Static method in Rattlesnake/MOOSE for time-dependent radiation transport approximations, (3) improved treatment of neutron physics representations within TREAT using OpenMC, and (4) steady state modeling of the minimum critical core of the Transient Reactor Test Facility (TREAT).

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

NASA Astrophysics Data System (ADS)

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

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

Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

NASA Astrophysics Data System (ADS)

Balooch, Mehdi; Olander, Donald R.; Terrani, Kurt A.; Hosemann, Peter; Casella, Andrew M.; Senor, David J.; Buck, Edgar C.

2017-04-01

A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.

Connor P. Nash | NREL

Science.gov Websites

catalysts Electrochemical reactor design and operation for transformation of biomass-derived intermediates Education B.S., Chemical and Biological Engineering, Colorado State University, 2010-2014 Professional

SMR Re-Scaling and Modeling for Load Following Studies

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

Hoover, K.; Wu, Q.; Bragg-Sitton, S.

2016-11-01

This study investigates the creation of a new set of scaling parameters for the Oregon State University Multi-Application Small Light Water Reactor (MASLWR) scaled thermal hydraulic test facility. As part of a study being undertaken by Idaho National Lab involving nuclear reactor load following characteristics, full power operations need to be simulated, and therefore properly scaled. Presented here is the scaling analysis and plans for RELAP5-3D simulation.

Reactor Experiments at the University of Minnesota.

DTIC Science & Technology

1987-07-15

metallurgy; zinc, zinc oxide; solar thermal,’ solar Pi% thermoelectrochemical’ water splitting, separation devices; reactors e, ? 20. AeSiRACT (Continue oe...reported. Water splitting, recovery of hydrogen 4. and sulfur from hydrogen sulfide, electrolysis of zinc oxide in vapor and liquid phases, oil...CH4-CO2 reforming process. 2. Hydrogen production from water and the production of hydrogen and sulfur (or ammonia and sulfuric acid) from H2S. 3

MTR MAIN FLOOR. MEN DEMONSTRATE INSERTION OF DUMMY PLUG INTO ...

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

MTR MAIN FLOOR. MEN DEMONSTRATE INSERTION OF DUMMY PLUG INTO AN MTR BEAM HOLE. ONE MAN CHECKS RADIATION LEVEL AT THE END OF THE UNIVERSAL COFFIN, WHILE ANOTHER USES TOOL TO INSERT PLUG INTO HOLE THROUGH COFFIN. MEN WEAR "ANTI-C" (ANTI-CONTAMINATION) CLOTHING. INL NEGATIVE NO. 6198. R.G. Larsen, Photographer, 6/27/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Li/sub 2/O microsphere fabrication. Monthly letter progress report No. 1 for the period ending May 31, 1977

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

Schluderberg, D C

1977-06-01

The purpose of this task is to establish process parameters for the fabrication of lithium oxide (Li/sub 2/O) microspheres having properties which as closely as possible approximate those required for the design characteristics of the University of Wisconsin design for a TOKAMAK-type fusion reactor utilizing the moving bed of Li/sub 2/O microspheres as both reactor coolant and tritium breeder.

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

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

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

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

Chemistry in microstructured reactors.

PubMed

Jähnisch, Klaus; Hessel, Volker; Löwe, Holger; Baerns, Manfred

2004-01-16

The application of microstructured reactors in the chemical process industry has gained significant importance in recent years. Companies that offer not only microstructured reactors, but also entire chemical process plants and services relating to them, are already in existence. In addition, many institutes and universities are active within this field, and process-engineering-oriented reviews and a specialized book are available. Microstructured systems can be applied with particular success in the investigation of highly exothermic and fast reactions. Often the presence of temperature-induced side reactions can be significantly reduced through isothermal operations. Although microstructured reaction techniques have been shown to optimize many synthetic procedures, they have not yet received the attention they deserve in organic chemistry. For this reason, this Review aims to address this by providing an overview of the chemistry in microstructured reactors, grouped into liquid-phase, gas-phase, and gas-liquid reactions.

Neutron flux and gamma dose measurement in the BNCT irradiation facility at the TRIGA reactor of the University of Pavia

NASA Astrophysics Data System (ADS)

Bortolussi, S.; Protti, N.; Ferrari, M.; Postuma, I.; Fatemi, S.; Prata, M.; Ballarini, F.; Carante, M. P.; Farias, R.; González, S. J.; Marrale, M.; Gallo, S.; Bartolotta, A.; Iacoviello, G.; Nigg, D.; Altieri, S.

2018-01-01

University of Pavia is equipped with a TRIGA Mark II research nuclear reactor, operating at a maximum steady state power of 250 kW. It has been used for many years to support Boron Neutron Capture Therapy (BNCT) research. An irradiation facility was constructed inside the thermal column of the reactor to produce a sufficient thermal neutron flux with low epithermal and fast neutron components, and low gamma dose. In this irradiation position, the liver of two patients affected by hepatic metastases from colon carcinoma were irradiated after borated drug administration. The facility is currently used for cell cultures and small animal irradiation. Measurements campaigns have been carried out, aimed at characterizing the neutron spectrum and the gamma dose component. The neutron spectrum has been measured by means of multifoil neutron activation spectrometry and a least squares unfolding algorithm; gamma dose was measured using alanine dosimeters. Results show that in a reference position the thermal neutron flux is (1.20 ± 0.03) ×1010 cm-2 s-1 when the reactor is working at the maximum power of 250 kW, with the epithermal and fast components, respectively, 2 and 3 orders of magnitude lower than the thermal component. The ratio of the gamma dose with respect to the thermal neutron fluence is 1.2 ×10-13 Gy/(n/cm2).

Microdosimetric evaluation of the neutron field for BNCT at Kyoto University reactor by using the PHITS code.

PubMed

Baba, H; Onizuka, Y; Nakao, M; Fukahori, M; Sato, T; Sakurai, Y; Tanaka, H; Endo, S

2011-02-01

In this study, microdosimetric energy distributions of secondary charged particles from the (10)B(n,α)(7)Li reaction in boron-neutron capture therapy (BNCT) field were calculated using the Particle and Heavy Ion Transport code System (PHITS). The PHITS simulation was performed to reproduce the geometrical set-up of an experiment that measured the microdosimetric energy distributions at the Kyoto University Reactor where two types of tissue-equivalent proportional counters were used, one with A-150 wall alone and another with a 50-ppm-boron-loaded A-150 wall. It was found that the PHITS code is a useful tool for the simulation of the energy deposited in tissue in BNCT based on the comparisons with experimental results.

78 FR 11688 - Notice of Issuance of Amendment to Facility License R-77 Incorporating a Decommissioning Plan for...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-19

... Reactor at the State University of New York at Buffalo AGENCY: Nuclear Regulatory Commission. ACTION... University of New York at Buffalo (UB) decommissioning plan (DP) by amendment to the Facility License R-77... in the NRC Library at http://www.nrc.gov/reading-rm/adams.html . To begin the search, select ``ADAMS...

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

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

Graslund, C.; Hellstrand, E.

Sweden benefits in many ways from the reactor safety research performed in other countries. Its own activity complements this effort, but a certain fraction is oriented toward safety issues that are intimately related to the special design of the ASEA-ATOM boiling-water reactor. Through the availability of the decommissioned Marviken reactor plant, Sweden has been able to play a leading role in integral containment experiments with international participation. Joint efforts with other countries are now devoted to defining new large-scale experiments to be performed in the unique Marviken facility. The largest portion of the safety research program in Sweden is performedmore » by Studsvik Energiteknik AB, but various universities, consultant firms, and research institutes are also involved. In addition, a substantial amount of work is done by the reactor vendor ASEA-ATOM. The overall annual budget is at present between $7 and $8 million, with three governmental authorities as the main financing bodies.« less

Investigation of Chirality Selection Mechanism of Single Walled Carbon Nanotube-3

DTIC Science & Technology

2017-12-14

however, several universal and intrinsic problems remain. First, since the dewetting of a thin catalyst film into particles upon heating is a... heated to 800 °C in 15 minutes under Ar atmosphere, maintained for various times, and cooled down to room temperature. - Annealing of Fe-implanted...located 12 cm downstream from the middle of the tube reactor. Then the reactor was heated to 820 °C over 15 min with flowing Ar gas. During the ramping

Development of a Short-Lived Radioisotope Production Service (SRPS) for CTTC at the University of Alberta SLOWPOKE Reactor Facility

DTIC Science & Technology

2004-12-01

abundance and neutron cross-section of 17~r (the precursor of 171 Er~~limit,tb~’Xi!Y o~_ 171Er that could be produced and delivered to Suf field to about...the radioisotope being produced. Additional factors relate to the irradiation conditions and include the reactor neutron flux and the irradiation...generally be desirable to have single radioisotopic sources with nuclear characteristics (e.g., half-life, gamma-ray energies and emission rates

Kinetics of thermophilic anaerobes in fixed-bed reactors.

PubMed

Perez, M; Romero, L I; Sales, D

2001-08-01

The main objective of this study is to estimate growth kinetic constants and the concentration of "active" attached biomass in two anaerobic thermophilic reactors which contain different initial sizes of immobilized anaerobic mixed cultures and decompose distillery wastewater. This paper studies the substrate decomposition in two lab-scale fixed-bed reactors operating at batch conditions with corrugated tubes as support media. It can be demonstrated that high micro-organisms-substrate ratios favor the degradation activity of the different anaerobic cultures, allowing the stable operation without lag-phases and giving better quality in effluent. The kinetic parameters obtained--maximum specific growth rates (mu(max)), non-biodegradable substrate (S(NB)) and "active or viable biomass" concentrations (X(V0))--were obtained by applying the Romero kinetic model [L.I. Romero, 1991. Desarrollo de un modelo matemático general para los procesos fermentativos, Cinética de la degradación anaerobia, Ph.D. Thesis, University of Cádiz (Spain), Serv. Pub. Univ. Cádiz], with COD as substrate and methane (CH4) as the main product of the anaerobic process. This method is suitable to calculate and to differentiate the main kinetic parameters of both the total anaerobic mixed culture and the methanogenic population. Comparison of experimental measured concentration of volatile attached solids (VS(att)) in both reactors with the estimated "active" biomass concentrations obtained by applying Romero kinetic model [L.I. Romero, 1991. Desarrollo de un modelo matemático general para los procesos fermentativos, Cinética de la degradación anaerobia, Ph.D. Thesis, University of Cádiz (Spain), Serv. Pub. Univ. Cádiz] shows that a large amount of inert matter is present in the fixed-bed reactor.

Design study of multi-imaging plate system for BNCT irradiation field at Kyoto university reactor.

PubMed

Tanaka, Kenichi; Sakurai, Yoshinori; Kajimoto, Tsuyoshi; Tanaka, Hiroki; Takata, Takushi; Endo, Satoru

2016-09-01

The converter configuration for a multi-imaging plate system was investigated for the application of quality assurance in the irradiation field profile for boron neutron capture therapy. This was performed by the simulation calculation using the PHITS code in the fields at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. The converter constituents investigated were carbon for gamma rays, and polyethylene with and without LiF at varied (6)Li concentration for thermal, epithermal, and fast neutrons. Consequently, potential combinations of the converters were found for two components, gamma rays and thermal neutrons, for the standard thermal neutron mode and three components of gamma rays, epithermal neutrons, and thermal or fast neutrons, for the standard mixed or epithermal neutron modes, respectively. Copyright © 2016 Elsevier Ltd. All rights reserved.

Modelling of the anti-neutrino production and spectra from a Magnox reactor

NASA Astrophysics Data System (ADS)

Mills, Robert W.; Mountford, David J.; Coleman, Jonathon P.; Metelko, Carl; Murdoch, Matthew; Schnellbach, Yan-Jie

2018-01-01

The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor operation was calculated. The results of the preliminary calculations are shown and limitations in the methods and data discussed.

Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

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

Heeger, Karsten M.

2014-09-13

This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zeromore » $$\\theta_{13}$$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.« less

A comparison of neutron and gamma damage effects on silica glass in a nuclear reactor radiation environment

NASA Astrophysics Data System (ADS)

Holcomb, David E.; Miller, Don W.

1993-08-01

A study of the relative damage effects of neutrons and gamma rays on silica glass in a nuclear reactor radiation environment is reported. The neutron and gamma energy spectra of the Ohio State University Research Reactor beam port #1 were applied to silica glass to obtain primary knock-on charged particle energy spectra. The resultant charged particle spectra were then applied to the polyatomic forms of the Lindhard et al. integrodifferential equation for damage energy and the Parkin and Coulter integrodifferential equation for net atomic displacement. The results show that near a nuclear reactor core the vast majority of the dose to silica is due to gamma rays (factor of roughly 40) and that neutrons cause much more displacement damage than gamma rays (35 times the oxygen displacement rate and 500 times the silicon displacement rate). However, pure silica core optical fibers irradiated in a nuclear reactor's mixed neutron/gamma environment exhibit little difference in transmission loss on an equal dose basis compared to fibers irradiated in a gamma only environment, indicating that atomic displacement is not a significant damage mechanism.

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

Scientists warn of 'trillion-dollar' spent-fuel risk

NASA Astrophysics Data System (ADS)

Gwynne, Peter

2016-07-01

A study by two Princeton University physicists suggests that a major fire in the spent nuclear fuel stored on the sites of US nuclear reactors could “dwarf the horrific consequences of the Fukushima accident”.

Argonne HEP Lunch Seminars

Science.gov Websites

reactor neutrino experiment July 29, 2008 August 5, 2008 Adam Clark (Rockhurst University) and Teng Jian August 12, 2008 Peter de Maagt THz Technology August 19, 2008 Razmick Mirzoyan Recent Developments on Low

Unique educational opportunities at the Missouri University research reactor

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

Ketring, A.R.; Ross, F.K.; Spate, V.

1997-12-01

Since the Missouri University Research Reactor (MURR) went critical in 1966, it has been a center where students from many departments conduct their graduate research. In the past three decades, hundreds of graduate students from the MU departments of chemistry, physics, anthropology, nuclear engineering, etc., have received masters and doctoral degrees based on research using neutrons produced at MURR. More recently, the educational opportunities at MURR have been expanded to include undergraduate students and local high school students. Since 1989 MURR has participated in the National Science Foundation-funded Research Experience for Undergraduates (REU) program. As part of this program, undergraduatemore » students from universities and colleges throughout the United States come to MURR and get hands-on research experience during the summer. Another program, started in 1994 by the Nuclear Analysis Program at MURR, allows students from a local high school to conduct a neutron activation analysis (NAA) experiment. We also conduct tours of the center, where we describe the research and educational programs at MURR to groups of elementary school children, high school science teachers, state legislators, professional organizations, and many other groups.« less

A Research Reactor Concept to Support NTP Development

NASA Technical Reports Server (NTRS)

Eades, Michael J.; Blue, T. E.; Gerrish, Harold P.; Hardin, Leroy A.

2014-01-01

In support of efforts for research into the design and development of man rated Nuclear Thermal Propulsion (NTP), the National Aeronautics and Space Administration (NASA), Marshall Space Flight Center (MSFC), is evaluating the potential for building a Nuclear Regulatory Commission (NRC) licensed NTP based research reactor (NTPRR). The proposed NTPRR would be licensed by NASA and operated jointly by NASA and university partners. The purpose of the NTPRR would be used to perform further research into the technologies and systems needed for a successful NTP project and promote nuclear training and education.

ATR National Scientific User Facility 2013 Annual Report

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

Ulrich, Julie A.; Robertson, Sarah

2015-03-01

This is the 2013 Annual Report for the Advanced Test Reactor National Scientific User Facility. This report includes information on university-run research projects along with a description of the program and the capabilities offered researchers.

Molten Salts for High Temperature Reactors: University of Wisconsin Molten Salt Corrosion and Flow Loop Experiments -- Issues Identified and Path Forward

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

Piyush Sabharwall; Matt Ebner; Manohar Sohal

2010-03-01

Considerable amount of work is going on regarding the development of high temperature liquid salts technology to meet future process needs of Next Generation Nuclear Plant. This report identifies the important characteristics and concerns of high temperature molten salts (with lesson learned at University of Wisconsin-Madison, Molten Salt Program) and provides some possible recommendation for future work

Convective cooling in a pool-type research reactor

NASA Astrophysics Data System (ADS)

Sipaun, Susan; Usman, Shoaib

2016-01-01

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

Various methods to improve heat transfer in exchangers

NASA Astrophysics Data System (ADS)

Pavel, Zitek; Vaclav, Valenta

2015-05-01

The University of West Bohemia in Pilsen (Department of Power System Engineering) is working on the selection of effective heat exchangers. Conventional shell and tube heat exchangers use simple segmental baffles. It can be replaced by helical baffles, which increase the heat transfer efficiency and reduce pressure losses. Their usage is demonstrated in the primary circuit of IV. generation MSR (Molten Salt Reactors). For high-temperature reactors we consider the use of compact desk heat exchangers, which are small, which allows the integral configuration of reactor. We design them from graphite composites, which allow up to 1000°C and are usable as exchangers: salt-salt or salt-acid (e.g. for the hydrogen production). In the paper there are shown thermo-physical properties of salts, material properties and principles of calculations.

Study of the Open Loop and Closed Loop Oscillator Techniques

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

Imel, George R.; Baker, Benjamin; Riley, Tony

This report presents the progress and completion of a five-year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques.The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this report we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign tomore » measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems.« less

Study of the open loop and closed loop oscillator techniques

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

Baker, Benjamin; Riley, Tony; Langbehn, Adam

This paper presents some aspects of a five year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques. The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this paper we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign tomore » measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems. (authors)« less

Application of nuclear pumped laser to an optical self-powered neutron detector

NASA Astrophysics Data System (ADS)

Yamanaka, N.; Takahashi, H.; Iguchi, T.; Nakazawa, M.; Kakuta, T.; Yamagishi, H.; Katagiri, M.

1996-05-01

A Nuclear Pumped Laser (NPL) using 3He/Ne/Ar gas mixture is investigated for a purpose of applying to an optical self-powered neutron detector. Reactor experiments and simulations on lasing mechanism have been made to estimate the best gas pressure and mixture ratios on the threshold input power density (or thermal neutron flux) in 3He/Ne/Ar mixture. Calculational results show that the best mixture pressure is 3He/Ne/Ar=2280/60/100 Torr and thermal neutron flux threshold 5×1012 n/cm2 sec, while the reactor experiments made in the research reactor ``YAYOI'' of the University of Tokyo and ``JRR-4'' of JAERI also demonstrate that excitational efficiency is maximized in a similar gas mixture predicted by the calculation.

Effect of airflow on biodrying of gardening wastes in reactors.

PubMed

Colomer-Mendoza, F J; Herrera-Prats, L; Robles-Martínez, F; Gallardo-Izquierdo, A; Piña-Guzmán, A B

2013-05-01

Biodrying consists of reducing moisture by using the heat from aerobic bio-degradation. The parameters that control the process are: aeration, temperature during the process, initial moisture of biowaste, and temperature and relative humidity of the input air. Lawn mowing and garden waste from the gardens of the University Jaume I, Castellón (Spain) were used as a substrate. Biodrying was performed in 10 reactors with known air volumes from 0.88 to 6.42 L/(min x kg dry weight). To promote aeration, 5 of the reactors had 15% of a bulking agent added. The experiment lasted 20 days. After the experiments it was found that the bulking agent led to greater weight loss. However, the increased airflow rate was not linearly proportional to the weight loss.

How much does a tokamak reactor cost?

NASA Astrophysics Data System (ADS)

Freidberg, J.; Cerfon, A.; Ballinger, S.; Barber, J.; Dogra, A.; McCarthy, W.; Milanese, L.; Mouratidis, T.; Redman, W.; Sandberg, A.; Segal, D.; Simpson, R.; Sorensen, C.; Zhou, M.

2017-10-01

The cost of a fusion reactor is of critical importance to its ultimate acceptability as a commercial source of electricity. While there are general rules of thumb for scaling both overnight cost and levelized cost of electricity the corresponding relations are not very accurate or universally agreed upon. We have carried out a series of scaling studies of tokamak reactor costs based on reasonably sophisticated plasma and engineering models. The analysis is largely analytic, requiring only a simple numerical code, thus allowing a very large number of designs. Importantly, the studies are aimed at plasma physicists rather than fusion engineers. The goals are to assess the pros and cons of steady state burning plasma experiments and reactors. One specific set of results discusses the benefits of higher magnetic fields, now possible because of the recent development of high T rare earth superconductors (REBCO); with this goal in mind, we calculate quantitative expressions, including both scaling and multiplicative constants, for cost and major radius as a function of central magnetic field.

Preliminary Tritium Management Design Activities at ORNL

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

Harrison, Thomas J.; Felde, David K.; Logsdon, Randall J.

2016-09-01

Interest in salt-cooled and salt-fueled reactors has increased over the last decade (Forsberg et al. 2016). Several private companies and universities in the United States, as well as governments in other countries, are developing salt reactor designs and/or technology. Two primary issues for the development and deployment of many salt reactor concepts are (1) the prevention of tritium generation and (2) the management of tritium to prevent release to the environment. In 2016, the US Department of Energy (DOE) initiated a research project under the Advanced Reactor Technology Program to (1) experimentally assess the feasibility of proposed methods for tritiummore » mitigation and (2) to perform an engineering demonstration of the most promising methods. This document describes results from the first year’s efforts to define, design, and build an experimental apparatus to test potential methods for tritium management. These efforts are focused on producing a final design document as the basis for the apparatus and its scheduled completion consistent with available budget and approvals for facility use.« less

NUCLEAR NEW BUILD-INTEGRATING CULTURAL DIFFERENCES IN RADIATION PROTECTION.

PubMed

Haemmerli, Valentin; Bryant, Peter A; Cole, Peter

2017-04-01

Across the world, we are seeing a resurgence in Nuclear New Build. In the UK alone, plans are under way for the construction of 10 new reactors, using 4 different reactor designs all of which are to be provided by foreign vendors, and operated by 3 newly formed licensees within the UK. As these new licensees embark on the task of establishing themselves and progressing the design and build of these reactors, there are challenges faced in integrating the Radiation Protection Requirements and Culture from the various Foreign Investors and Vendors into the UK 'Context'. The following paper identifies the origin of the Radiation Protection Requirements within the UK and foreign investor/vendor countries, in an attempt to integrate them into the UK licensing and approval process. Thus, allowing due credit to be taken for the regulatory regime of the foreign countries where these reactors originate. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Assessing the degree of plug flow in oxidation flow reactors (OFRs): a study on a potential aerosol mass (PAM) reactor

NASA Astrophysics Data System (ADS)

Mitroo, Dhruv; Sun, Yujian; Combest, Daniel P.; Kumar, Purushottam; Williams, Brent J.

2018-03-01

Oxidation flow reactors (OFRs) have been developed to achieve high degrees of oxidant exposures over relatively short space times (defined as the ratio of reactor volume to the volumetric flow rate). While, due to their increased use, attention has been paid to their ability to replicate realistic tropospheric reactions by modeling the chemistry inside the reactor, there is a desire to customize flow patterns. This work demonstrates the importance of decoupling tracer signal of the reactor from that of the tubing when experimentally obtaining these flow patterns. We modeled the residence time distributions (RTDs) inside the Washington University Potential Aerosol Mass (WU-PAM) reactor, an OFR, for a simple set of configurations by applying the tank-in-series (TIS) model, a one-parameter model, to a deconvolution algorithm. The value of the parameter, N, is close to unity for every case except one having the highest space time. Combined, the results suggest that volumetric flow rate affects mixing patterns more than use of our internals. We selected results from the simplest case, at 78 s space time with one inlet and one outlet, absent of baffles and spargers, and compared the experimental F curve to that of a computational fluid dynamics (CFD) simulation. The F curves, which represent the cumulative time spent in the reactor by flowing material, match reasonably well. We value that the use of a small aspect ratio reactor such as the WU-PAM reduces wall interactions; however sudden apertures introduce disturbances in the flow, and suggest applying the methodology of tracer testing described in this work to investigate RTDs in OFRs to observe the effect of modified inlets, outlets and use of internals prior to application (e.g., field deployment vs. laboratory study).

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

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

Kempf, F.J.

1964-12-10

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

Current status of nuclear engineering education

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

Palladino, N.J.

1975-09-01

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

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

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

Gerald E. Tripard; Keith G. Fox

2006-08-24

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

Laser Assisted CVD Growth of A1N and GaN

DTIC Science & Technology

1990-08-31

additional cost sharing. RESEARCH FACILITIES The york is being performed in the Howard University Laser Laboratory. This is a free-standing buildinq...would be used to optimize computer models of the laser induced CVD reactor. FACILITIES AND EQUIPMENT - ADDITIONAL COST SHARING This year Howard ... University has provided $45,000 for the purchase of an excimer laser to be shared by Dr. Crye for the diode laser probe experiments and another Assistant

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

Algora, A.; Rice, S.; Guadilla, V.

Preliminary results from beta decay studies of nuclei that are important for reactor applications are presented. The beta decays have been studied using the total absorption technique (TAS) and the pure beams provided by the JYFLTRAP system at the IGISOL facility of the University of Jyväskylä.

The Experimental Determination of Thermal Neutron Flux in the Radiochemistry Curriculum

ERIC Educational Resources Information Center

Grant, Patrick M.

1977-01-01

Describes an experiment for determining the thermal neutron flux of the light-water nuclear reactor at the University of California, Irvine. The difficulty of the activity can be varied to match the student's level of proficiency. (SL)

Twenty years of experience in monitoring 41Ar in a research reactor and decrease of its discharge into the environment.

PubMed

Fukui, M

2004-04-01

The radioactive gas 41Ar has been produced at high concentration by neutron activation near the reactor core in the Kyoto University Research Reactor. A pipe line for an exhaust stream, so-called sweep gas, was fabricated at the construction of the reactor in 1964 in order to exhale 41Ar from the facilities above to the environment. Other exhaust lines with decay tanks were established separately from the sweep line for both the cold neutron source in 1986 and the heavy-water tank in 1996, respectively, because a higher amount of 41Ar was thought to be produced from these facilities due to the improvement. As a result, a slight change in the flow rate of the exhaust was found to have a great deal of influence on both the 41Ar concentration in the reactor room and the rate of emission from the stack. By monitoring the exhaust air from the decay tanks, the mechanism for decreasing the emission was clarified together with identifying an obstacle, i.e., the condensate against the steady state flow, formed in the exhaust pipe. By setting the flow rate suitably in the exhaust line, the rate of 41Ar emission from the biological shielding into both the work place in the reactor room and the environment has been controlled as low as reasonably achievable.

Convective cooling in a pool-type research reactor

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

Sipaun, Susan, E-mail: susan@nm.gov.my; Usman, Shoaib, E-mail: usmans@mst.edu

2016-01-22

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to passmore » through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.« less

Neutron spectrometry and dosimetry study at two research nuclear reactors using Bonner sphere spectrometer (BSS), rotational spectrometer (ROSPEC) and cylindrical nested neutron spectrometer (NNS).

PubMed

Atanackovic, J; Matysiak, W; Hakmana Witharana, S S; Aslam, I; Dubeau, J; Waker, A J

2013-01-01

Neutron spectrometry and subsequent dosimetry measurements were undertaken at the McMaster Nuclear Reactor (MNR) and AECL Chalk River National Research Universal (NRU) Reactor. The instruments used were a Bonner sphere spectrometer (BSS), a cylindrical nested neutron spectrometer (NNS) and a commercially available rotational proton recoil spectrometer. The purposes of these measurements were to: (1) compare the results obtained by three different neutron measuring instruments and (2) quantify neutron fields of interest. The results showed vastly different neutron spectral shapes for the two different reactors. This is not surprising, considering the type of the reactors and the locations where the measurements were performed. MNR is a heavily shielded light water moderated reactor, while NRU is a heavy water moderated reactor. The measurements at MNR were taken at the base of the reactor pool, where a large amount of water and concrete shielding is present, while measurements at NRU were taken at the top of the reactor (TOR) plate, where there is only heavy water and steel between the reactor core and the measuring instrument. As a result, a large component of the thermal neutron fluence was measured at MNR, while a negligible amount of thermal neutrons was measured at NRU. The neutron ambient dose rates at NRU TOR were measured to be between 0.03 and 0.06 mSv h⁻¹, while at MNR, these values were between 0.07 and 2.8 mSv h⁻¹ inside the beam port and <0.2 mSv h⁻¹ between two operating beam ports. The conservative uncertainty of these values is 15 %. The conservative uncertainty of the measured integral neutron fluence is 5 %. It was also found that BSS over-responded slightly due to a non-calibrated response matrix.

JPRS Report Science and Technology, Japan: Atomic Energy Society 1989 Annual Meeting.

DTIC Science & Technology

1989-10-13

Control Rod Hole in VHTRC-1 Core [F, Akino, T, Yamane, et al.] ,,, 5 Measurement of MEU [Medium Enriched Uranium ] Fuel Element Characteristics in...K. Yoshida, K. Kobayashi, I. Kimura , C. Yamanaka, and S. Nakai, Laser Laboratory,, Osaka University. Nuclear Reactor Laboratory, Kyoto University...1 core loaded with 278 fuel rods (4 percent enriched uranium ). The PNS target was placed at the back center of the 1/2 assembly on the fixed side

Elastic and inelastic neutron scattering cross sections for fission reactor applications

NASA Astrophysics Data System (ADS)

Hicks, S. F.; Chakraborty, A.; Combs, B.; Crider, B. P.; Downes, L.; Girgis, J.; Kersting, L. J.; Kumar, A.; Lueck, C. J.; McDonough, P. J.; McEllistrem, M. T.; Peters, E. E.; Prados-Estevz, F. M.; Schniederjan, J.; Sidwell, L.; Sigillito, A. J.; Vanhoy, J. R.; Watts, D.; Yates, S. W.

2013-04-01

Nuclear data important for the design and development of the next generation of light-water reactors and future fast reactors include neutron elastic and inelastic scattering cross sections on important structural materials, such as Fe, and on coolant materials, such as Na. These reaction probabilities are needed since neutron reactions impact fuel performance during irradiations and the overall efficiency of reactors. While neutron scattering cross sections from these materials are available for certain incident neutron energies, the fast neutron region, particularly above 2 MeV, has large gaps for which no measurements exist, or the existing uncertainties are large. Measurements have been made at the University of Kentucky Accelerator Laboratory to measure neutron scattering cross sections on both Fe and Na in the region where these gaps occur and to reduce the uncertainties on scattering from the ground state and first excited state of these nuclei. Results from measurements on Fe at incident neutron energies between 2 and 4 MeV will be presented and comparisons will be made to model calculations available from data evaluators.

Total absorption spectroscopy of fission fragments relevant for reactor antineutrino spectra

DOE PAGES

Fallot, M.; Porta, A.; Meur, L. Le; ...

2017-09-13

Here, the accurate determination of reactor antineutrino spectra remains a very active research topic for which new methods of study have emerged in recent years. Indeed, following the long-recognized reactor anomaly (measured antineutrino deficit in short baseline reactor experiments when compared with spectral predictions), the three international reactor neutrino experiments Double Chooz, Daya Bay and Reno have recently demonstrated the existence of spectral distortions in their measurements with respect to the same predictions. These spectral predictions were obtained through the conversion of integral beta-energy spectra obtained at the ILL research reactor. Several studies have shown that the underlying nuclear physicsmore » required for the conversion of these spectra into antineutrino spectra is not totally understood. An alternative to such converted spectra is a complementary approach that consists of determining the antineutrino spectrum by means of the measurement and processing of nuclear data. The beta properties of some key fission products suffer from the pandemonium effect which can be circumvented by the use of the Total Absorption Gamma-ray Spectroscopy technique (TAGS). The two main contributors to the Pressurized Water Reactor antineutrino spectrum in the region where the spectral distortion has been observed are 92Rb and 142Cs, which have been measured at the radioactive beam facility of the University of Jyvaskyla in two TAGS experiments. We present the results of the analysis of the TAGS measurements of the β-decay properties of 92Rb along with preliminary results on 142Cs and report on the measurements already performed.« less

Total absorption spectroscopy of fission fragments relevant for reactor antineutrino spectra

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

Fallot, M.; Porta, A.; Meur, L. Le

Here, the accurate determination of reactor antineutrino spectra remains a very active research topic for which new methods of study have emerged in recent years. Indeed, following the long-recognized reactor anomaly (measured antineutrino deficit in short baseline reactor experiments when compared with spectral predictions), the three international reactor neutrino experiments Double Chooz, Daya Bay and Reno have recently demonstrated the existence of spectral distortions in their measurements with respect to the same predictions. These spectral predictions were obtained through the conversion of integral beta-energy spectra obtained at the ILL research reactor. Several studies have shown that the underlying nuclear physicsmore » required for the conversion of these spectra into antineutrino spectra is not totally understood. An alternative to such converted spectra is a complementary approach that consists of determining the antineutrino spectrum by means of the measurement and processing of nuclear data. The beta properties of some key fission products suffer from the pandemonium effect which can be circumvented by the use of the Total Absorption Gamma-ray Spectroscopy technique (TAGS). The two main contributors to the Pressurized Water Reactor antineutrino spectrum in the region where the spectral distortion has been observed are 92Rb and 142Cs, which have been measured at the radioactive beam facility of the University of Jyvaskyla in two TAGS experiments. We present the results of the analysis of the TAGS measurements of the β-decay properties of 92Rb along with preliminary results on 142Cs and report on the measurements already performed.« less

Fabrication and testing of low-temperature catalytically active washcoat materials for next-generation vehicle catalytic converters.

DOT National Transportation Integrated Search

2014-01-01

With funds from this grant, Youngstown State University has purchased a : Quantachrome iQ Chemisorption Analyzer and built a Chemical Reactor with Gas : Chromatographer for catalyst characterization. PI has been involved in the purchase and : install...

Neutron Scattering Facilities

Science.gov Websites

Low Energy Neutron Source (LENS), Indiana University Cyclotron Facility, USA McMaster Nuclear Reactor Research, Gaithersburg, Maryland, USA Peruvian Institute of Nuclear Energy (IPEN), Lima, Peru Spallation Nuclear Science and Technology Organisation, Lucas Heights, Australia High-flux Advanced Neutron

Behavior of U 3Si 2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

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

Gamble, Kyle Allan Lawrence; Hales, Jason Dean; Barani, Tommaso

2016-09-01

As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Losmore » Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.« less

Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

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

A K Wertsching

2012-09-01

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

Accident analysis of heavy water cooled thorium breeder reactor

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

Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

2015-04-16

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k,more » and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.« less

Accident analysis of heavy water cooled thorium breeder reactor

NASA Astrophysics Data System (ADS)

Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

2015-04-01

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

Modeling of water lighting process and calculation of the reactor-clarifier to improve energy efficiency

NASA Astrophysics Data System (ADS)

Skolubovich, Yuriy; Skolubovich, Aleksandr; Voitov, Evgeniy; Soppa, Mikhail; Chirkunov, Yuriy

2017-10-01

The article considers the current questions of technological modeling and calculation of the new facility for cleaning natural waters, the clarifier reactor for the optimal operating mode, which was developed in Novosibirsk State University of Architecture and Civil Engineering (SibSTRIN). A calculation technique based on well-known dependences of hydraulics is presented. A calculation example of a structure on experimental data is considered. The maximum possible rate of ascending flow of purified water was determined, based on the 24 hour clarification cycle. The fractional composition of the contact mass was determined with minimal expansion of contact mass layer, which ensured the elimination of stagnant zones. The clarification cycle duration was clarified by the parameters of technological modeling by recalculating maximum possible upward flow rate of clarified water. The thickness of the contact mass layer was determined. Likewise, clarification reactors can be calculated for any other lightening conditions.

A review of carbide fuel corrosion for nuclear thermal propulsion applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1993-10-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1994-07-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

Soil bed reactor work of the Environmental Research Lab. of the University of Arizona in support of the research and development of Biosphere 2

NASA Technical Reports Server (NTRS)

Frye, Robert

1990-01-01

Research at the Environmental Research Lab in support of Biosphere 2 was both basic and applied in nature. One aspect of the applied research involved the use of biological reactors for the scrubbing of trace atmospheric organic contaminants. The research involved a quantitative study of the efficiency of operation of Soil Bed Reactors (SBR) and the optimal operating conditions for contaminant removal. The basic configuration of a SBR is that air is moved through a living soil that supports a population of plants. Upon exposure to the soil, contaminants are either passively adsorbed onto the surface of soil particles, chemically transformed in the soil to usable compounds that are taken up by the plants or microbes as a metabolic energy source and converted to CO2 and water.

75 FR 36710 - The Texas Engineering Experiment Station/Texas A&M University System; Notice of Acceptance for...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-28

... for the Nuclear Science Center Reactor and Order Imposing Procedures for Access To Safeguards Information and Sensitive Unclassified Non- Safeguards Information AGENCY: Nuclear Regulatory Commission. ACTION: Notice of acceptance for docketing. FOR FURTHER INFORMATION CONTACT: Christian Cowdrey, Project...

77 FR 13376 - Notice of License Termination for the University of Arizona Research Reactor, License No. R-52

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-06

..., alpha plus beta measurements, smear sampling, and soil sampling activities. As a result of the...-Agency Radiation Survey and Site Investigation Manual.'' The U of AZ submitted a revised FSS Plan on... and that the [[Page 13377

Nuclear Education and Training Programs of Potential Interest to Utilities.

ERIC Educational Resources Information Center

Atomic Energy Commission, Washington, DC.

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

Coupling Schemes for Multiphysics Reactor Simulation

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

Vijay Mahadeven; Jean Ragusa

2007-11-01

This report documents the progress of the student Vijay S. Mahadevan from the Nuclear Engineering Department of Texas A&M University over the summer of 2007 during his visit to the INL. The purpose of his visit was to investigate the physics-based preconditioned Jacobian-free Newton-Krylov method applied to physics relevant to nuclear reactor simulation. To this end he studied two test problems that represented reaction-diffusion and advection-reaction. These two test problems will provide the basis for future work in which neutron diffusion, nonlinear heat conduction, and a twophase flow model will be tightly coupled to provide an accurate model of amore » BWR core.« less

Environmental Monitoring and Characterization of Radiation Sources on UF Campus Using a Large Volume NaI Detector

NASA Astrophysics Data System (ADS)

Bruner, Jesse A.; Gardiner, Hannah E.; Jordan, Kelly A.; Baciak, James E.

2016-09-01

Environmental radiation surveys are important for applications such as safety and regulations. This is especially true for areas exposed to emissions from nuclear reactors, such as the University of Florida Training Reactor (UFTR). At the University of Florida, surveys are performed using the RSX-1 NaI detector, developed by Radiation Solutions Inc. The detector uses incoming gamma rays and an Advanced Digital Spectrometer module to produce a linear energy spectrum. These spectra can then be analyzed in real time with a personal computer using the built in software, RadAssist. We report on radiation levels around the University of Florida campus using two mobile detection platforms, car-borne and cart-borne. The car-borne surveys provide a larger, broader map of campus radiation levels. On the other hand, cart-borne surveys provide a more detailed radiation map because of its ability to reach places on campus cars cannot go. Throughout the survey data, there are consistent radon decay product energy peaks in addition to other sources such as medical I-131 found in a large crowd of people. Finally, we investigate further applications of this mobile detection platform, such as tracking the Ar-41 plume emitted from the UFTR and detection of potential environmental hazards.

Nitrous oxide from moving bed based integrated fixed film activated sludge membrane bioreactors.

PubMed

Mannina, Giorgio; Capodici, Marco; Cosenza, Alida; Di Trapani, Daniele; Laudicina, Vito Armando; Ødegaard, Hallvard

2017-02-01

The present paper reports the results of a nitrous oxide (N 2 O) production investigation in a moving bed based integrated fixed film activated sludge (IFAS) membrane bioreactor (MBR) pilot plant designed in accordance with the University of Cape Town layout for biological phosphorous removal. Gaseous and liquid samples were collected in order to measure the gaseous as well as the dissolved concentration of N 2 O. Furthermore, the gas flow rate from each reactor was measured and the gas flux was estimated. The results confirmed that the anoxic reactor represents the main source of nitrous oxide production. A significant production of N 2 O was, however, also found in the anaerobic reactor, thus indicating a probable occurrence of the denitrifying phosphate accumulating organism activity. The highest N 2 O fluxes were emitted from the aerated reactors (3.09 g N 2 ON m -2  h -1 and 9.87 g N 2 ON m -2  h -1 , aerobic and MBR tank, respectively). The emission factor highlighted that only 1% of the total treated nitrogen was emitted from the pilot plant. Furthermore, the measured N 2 O concentrations in the permeate flow were comparable with other reactors. Nitrous oxide mass balances outlined a moderate production also in the MBR reactor despite the low hydraulic retention time. On the other hand, the mass balance showed that in the aerobic reactor a constant consumption of nitrous oxide (up to almost 15 mg N 2 O h -1 ) took place, due to the high amount of stripped gas. Copyright © 2016 Elsevier Ltd. All rights reserved.

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

Soldatov, Alexey I.

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

Radioactive and other environmental threats to the United States and the Arctic resulting from past Soviet activities

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

NONE

Earlier this year the Senate Intelligence Committee began to receive reports from environmental and nuclear scientists in Russia detailing the reckless nuclear waste disposal practices, nuclear accidents and the use of nuclear detonations. We found that information disturbing to say the least. Also troubling is the fact that 15 Chernobyl style RBMK nuclear power reactors continue to operate in the former Soviet Union today. These reactors lack a containment structure and they`re designed in such a way that nuclear reaction can actually increase when the reactor overheats. As scientists here at the University of Alaska have documented, polar air massesmore » and prevailing weather patterns provide a pathway for radioactive contaminants from Eastern Europe and Western Russia, where many of these reactors are located. The threats presented by those potential radioactive risks are just a part of a larger Arctic pollution problem. Every day, industrial activities of the former Soviet Union continue to create pollutants. I think we should face up to the reality that in a country struggling for economic survival, environment protection isn`t necessarily the high priority. And that could be very troubling news for the Arctic in the future.« less

Chemistry and radiochemistry of As, Re and Rh isotopes relevant to radiopharmaceutical applications: high specific activity radionuclides for imaging and treatment.

PubMed

Feng, Yutian; Phelps, Tim E; Carroll, Valerie; Gallazzi, Fabio; Sieckman, Gary; Hoffman, Timothy J; Barnes, Charles L; Ketring, Alan R; Hennkens, Heather M; Jurisson, Silvia S

2017-10-31

The chemistry and radiochemistry of high specific activity radioisotopes of arsenic, rhenium and rhodium are reviewed with emphasis on University of Missouri activities over the past several decades, and includes recent results. The nuclear facilities at the University of Missouri (10 MW research reactor and 16.5 MeV GE PETtrace cyclotron) allow research and development into novel theranostic radionuclides. The production, separation, enriched target recovery, radiochemistry, and chelation chemistry of 72,77 As, 186,188 Re and 105 Rh are discussed.

Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

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

Bess, J. D.; Briggs, J. B.; Gulliford, J.

Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energymore » Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning is the critical experiments with fast reactor fuel rods in water, interesting in terms of justification of nuclear safety during transportation and storage of fresh and spent fuel. These reports provide a detailed review of the experiment, designate the area of their application and include results of calculations on modern systems of constants in comparison with the estimated experimental data.« less

Non-Destructive Analysis of Natural Uranium Pellet

NASA Astrophysics Data System (ADS)

Wigley, Samantha; Glennon, Kevin; Kitcher, Evans; Folden, Cody

2017-09-01

As part of ongoing nuclear forensics research, samples of natUO2 have been irradiated in a thermal neutron spectrum at the University of Missouri Research Reactor (MURR) with the goal of simulating a pressurized heavy water reactor. Non-destructive gamma ray analysis has been performed on the samples to assay various nuclides in order to determine the burnup and time since irradiation. The quantity of 137Cs was used to determine the burnup directly, and a maximum likelihood method has been used to estimate both the burnup and the time since irradiation. This poster will discuss the most recent results of these analyses. National Science Foundation (PHY-1659847), Department of Energy (DE-FG02-93ER40773).

Calculating Pi Using the Monte Carlo Method

ERIC Educational Resources Information Center

Williamson, Timothy

2013-01-01

During the summer of 2012, I had the opportunity to participate in a research experience for teachers at the center for sustainable energy at Notre Dame University (RET @ cSEND) working with Professor John LoSecco on the problem of using antineutrino detection to accurately determine the fuel makeup and operating power of nuclear reactors. During…

76 FR 60091 - Notice of Availability of Environmental Assessment and Finding of No Significant Impact for the...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-28

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-407; NRC-2011-0153] Notice of Availability of Environmental Assessment and Finding of No Significant Impact for the University of Utah Nuclear Reactor Facility; Facility Operating License No. R-126 AGENCY: Nuclear Regulatory Commission. ACTION: Notice of...

(US-Japan workshop on fusion neutronics at Osaka University, Osaka, Japan, May 25--26, and visits to Savar, Bangladesh, April 20--May 19, Beijing, China, May 21--24, and Tokai, Japan, May 29--30): Foreign trip report

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

White, J.E.

1989-06-15

This trip was initiated by a request from IAEA for USA expert assistance in Bangladesh. The Bangladesh Atomic Energy Commission (BAEC) had acquired a 3MW TRIGA MARK-II research reactor and their specialist needed help in the development of computer codes and data for the effective utilization and analysis of their reactor. Nuclear data recognized as an international standard was installed on a BAEC computer at Savar. The traveler was invited to China (PRC) to discuss multigroup cross section processing methods used at ORNL. Also, the traveler participated in a US-Japan workshop on fusion neutronics at Osaka University where he discussedmore » the activities of RSIC. An orientation visit to JAERI resulted in the collection of information of potential use in US DOE programs, a better understanding of their research activities, and an opportunity to develop personal relationships with JAERI staff.« less

Development of a repetitive compact torus injector

NASA Astrophysics Data System (ADS)

Onchi, Takumi; McColl, David; Dreval, Mykola; Rohollahi, Akbar; Xiao, Chijin; Hirose, Akira; Zushi, Hideki

2013-10-01

A system for Repetitive Compact Torus Injection (RCTI) has been developed at the University of Saskatchewan. CTI is a promising fuelling technology to directly fuel the core region of tokamak reactors. In addition to fuelling, CTI has also the potential for (a) optimization of density profile and thus bootstrap current and (b) momentum injection. For steady-state reactor operation, RCTI is necessary. The approach to RCTI is to charge a storage capacitor bank with a large capacitance and quickly charge the CT capacitor bank through a stack of integrated-gate bipolar transistors (IGBTs). When the CT bank is fully charged, the IGBT stack will be turned off to isolate banks, and CT formation/acceleration sequence will start. After formation of each CT, the fast bank will be replenished and a new CT will be formed and accelerated. Circuits for the formation and the acceleration in University of Saskatchewan CT Injector (USCTI) have been modified. Three CT shots at 10 Hz or eight shots at 1.7 Hz have been achieved. This work has been sponsored by the CRC and NSERC, Canada.

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

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

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

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

Dependency of the Reynolds number on the water flow through the perforated tube

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

Závodný, Zdenko, E-mail: zdenko.zavodny@stuba.sk; Bereznai, Jozef, E-mail: jozef.bereznai@stuba.sk; Urban, František

Safe and effective loading of nuclear reactor fuel assemblies demands qualitative and quantitative analysis of the relationship between the coolant temperature in the fuel assembly outlet, measured by the thermocouple, and the mean coolant temperature profile in the thermocouple plane position. It is not possible to perform the analysis directly in the reactor, so it is carried out using measurements on the physical model, and the CFD fuel assembly coolant flow models. The CFD models have to be verified and validated in line with the temperature and velocity profile obtained from the measurements of the cooling water flowing in themore » physical model of the fuel assembly. Simplified physical model with perforated central tube and its validated CFD model serve to design of the second physical model of the fuel assembly of the nuclear reactor VVER 440. Physical model will be manufactured and installed in the laboratory of the Institute of Energy Machines, Faculty of Mechanical Engineering of the Slovak University of Technology in Bratislava.« 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

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

Frybort, Jan

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

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

NASA Astrophysics Data System (ADS)

Esen, Ayse Nur; Haciyakupoglu, Sevilay

2016-02-01

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

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

Metz, Lori A.; Friese, Judah I.; Finn, Erin C.

Critical assemblies provide one method of achieving a fast neutron spectrum that is close to a 235U fission-energy neutron spectrum for nuclear data measurements. Previous work has demonstrated the use of a natural boron carbide capsule for spectral-tailoring in a mixed spectrum reactor as an alternate and complementary method for performing fission-energy neutron experiments. Previous fission products measurements showed that the neutron spectrum achievable with natural boron carbide was not as hard as what can be achieved with critical assemblies. New measurements performed with the Washington State University TRIGA reactor using a boron carbide capsule 96% enriched in 10B formore » irradiations resulted in a neutron spectrum very similar to a critical assembly and a pure 235U fission spectrum. The current work describes an experiment involving a highly-enriched uranium target irradiated under the new 10B4C capsule. Fission product yields were measured following radiochemical separations and are presented here. Reactor dosimetry measurements for characterizing neutron spectra and fluence for the enriched boron carbide capsule and critical assemblies are also discussed.« less

Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

NASA Astrophysics Data System (ADS)

Doshi, Bharat; Reddy, D. Chenna

2017-04-01

Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014

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

Ogden, Dan

Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014 Highlights • Rory Kennedy, Dan Ogden and Brenden Heidrich traveled to Germantown October 6-7, for a review of the Infrastructure Management mission with Shane Johnson, Mike Worley, Bradley Williams and Alison Hahn from NE-4 and Mary McCune from NE-3. Heidrich briefed the group on the project progress from July to October 2014 as well as the planned path forward for FY15. • Jim Cole gave two invited university seminars at Ohio State University and University of Florida, providing an overview of NSUF including available capabilities and themore » process for accessing facilities through the peer reviewed proposal process. • Jim Cole and Rory Kennedy co-chaired the NuMat meeting with Todd Allen. The meeting, sponsored by Elsevier publishing, was held in Clearwater, Florida, and is considered one of the premier nuclear fuels and materials conferences. Over 340 delegates attended with 160 oral and over 200 posters presented over 4 days. • Thirty-one pre-applications were submitted for NSUF access through the NE-4 Combined Innovative Nuclear Research Funding Opportunity Announcement. • Fourteen proposals were received for the NSUF Rapid Turnaround Experiment Summer 2014 call. Proposal evaluations are underway. • John Jackson and Rory Kennedy attended the Nuclear Fuels Industry Research meeting. Jackson presented an overview of ongoing NSUF industry research.« less

Dismantling the nuclear research reactor Thetis

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

Michiels, P.

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

Modeling of the HiPco process for carbon nanotube production. II. Reactor-scale analysis

NASA Technical Reports Server (NTRS)

Gokcen, Tahir; Dateo, Christopher E.; Meyyappan, M.

2002-01-01

The high-pressure carbon monoxide (HiPco) process, developed at Rice University, has been reported to produce single-walled carbon nanotubes from gas-phase reactions of iron carbonyl in carbon monoxide at high pressures (10-100 atm). Computational modeling is used here to develop an understanding of the HiPco process. A detailed kinetic model of the HiPco process that includes of the precursor, decomposition metal cluster formation and growth, and carbon nanotube growth was developed in the previous article (Part I). Decomposition of precursor molecules is necessary to initiate metal cluster formation. The metal clusters serve as catalysts for carbon nanotube growth. The diameter of metal clusters and number of atoms in these clusters are some of the essential information for predicting carbon nanotube formation and growth, which is then modeled by the Boudouard reaction with metal catalysts. Based on the detailed model simulations, a reduced kinetic model was also developed in Part I for use in reactor-scale flowfield calculations. Here this reduced kinetic model is integrated with a two-dimensional axisymmetric reactor flow model to predict reactor performance. Carbon nanotube growth is examined with respect to several process variables (peripheral jet temperature, reactor pressure, and Fe(CO)5 concentration) with the use of the axisymmetric model, and the computed results are compared with existing experimental data. The model yields most of the qualitative trends observed in the experiments and helps to understanding the fundamental processes in HiPco carbon nanotube production.

Nuclear space power safety and facility guidelines study

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

Mehlman, W.F.

1995-09-11

This report addresses safety guidelines for space nuclear reactor power missions and was prepared by The Johns Hopkins University Applied Physics Laboratory (JHU/APL) under a Department of Energy grant, DE-FG01-94NE32180 dated 27 September 1994. This grant was based on a proposal submitted by the JHU/APL in response to an {open_quotes}Invitation for Proposals Designed to Support Federal Agencies and Commercial Interests in Meeting Special Power and Propulsion Needs for Future Space Missions{close_quotes}. The United States has not launched a nuclear reactor since SNAP 10A in April 1965 although many Radioisotope Thermoelectric Generators (RTGs) have been launched. An RTG powered system ismore » planned for launch as part of the Cassini mission to Saturn in 1997. Recently the Ballistic Missile Defense Office (BMDO) sponsored the Nuclear Electric Propulsion Space Test Program (NEPSTP) which was to demonstrate and evaluate the Russian-built TOPAZ II nuclear reactor as a power source in space. As of late 1993 the flight portion of this program was canceled but work to investigate the attributes of the reactor were continued but at a reduced level. While the future of space nuclear power systems is uncertain there are potential space missions which would require space nuclear power systems. The differences between space nuclear power systems and RTG devices are sufficient that safety and facility requirements warrant a review in the context of the unique features of a space nuclear reactor power system.« less

On the Use of Interactive Texts in Undergraduate Chemical Reaction Engineering Courses: A Pedagogical Experience

ERIC Educational Resources Information Center

Asensio, Daniela A.; Barassi, Francisca J.; Zambon, Mariana T.; Mazza, Germán D.

2010-01-01

This paper describes the results of a pedagogical experience carried out at the University of Comahue, Argentina, with an interactive text (IT) concerning Homogeneous Chemical Reactors Analysis. The IT was built on the frame of the "Mathematica" software with the aim of providing students with a robust computational tool. Students'…

Developments of Spent Nuclear Fuel Pyroprocessing Technology at Idaho National Laboratory

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

Michael F. Simpson

This paper summarizes research in used fuel pyroprocessing that has been published by Idaho National Laboratory over the last decade. It includes work done both on treatment of Experimental Breeder Reactor-II and development of advanced technology for potential scale-up and commercialization. Collaborations with universities and other laboratories is included in the cited work.

75 FR 47646 - Idaho State University; Notice of Issuance of Director's Decision

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-06

... detector from the reactor core. In the petition Dr. Crawford states that this was cited as an Apparent... detector. The personnel roof access hatch was also addressed in Rev. 3 and Rev. 4 of the Physical Security... procedures to wear protective clothing to routinely remove the activated startup channel detector from the...

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

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

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

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

DOE PAGES

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

2017-03-26

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

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.

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.

Reduce, reuse and recycle: a green solution to Canada's medical isotope shortage.

PubMed

Galea, R; Ross, C; Wells, R G

2014-05-01

Due to the unforeseen maintenance issues at the National Research Universal (NRU) reactor at Chalk River and coincidental shutdowns of other international reactors, a global shortage of medical isotopes (in particular technetium-99m, Tc-99m) occurred in 2009. The operation of these research reactors is expensive, their age creates concerns about their continued maintenance and the process results in a large amount of long-lived nuclear waste, whose storage cost has been subsidized by governments. While the NRU has since revived its operations, it is scheduled to cease isotope production in 2016. The Canadian government created the Non-reactor based medical Isotope Supply Program (NISP) to promote research into alternative methods for producing medical isotopes. The NRC was a member of a collaboration looking into the use of electron linear accelerators (LINAC) to produce molybdenum-99 (Mo-99), the parent isotope of Tc-99m. This paper outlines NRC's involvement in every step of this process, from the production, chemical processing, recycling and preliminary animal studies to demonstrate the equivalence of LINAC Tc-99m with the existing supply. This process stems from reusing an old idea, reduces the nuclear waste to virtually zero and recycles material to create a green solution to Canada's medical isotope shortage. © 2013 Published by Elsevier Ltd.

Continuous lactic acid fermentation using a plastic composite support biofilm reactor.

PubMed

Cotton, J C; Pometto, A L; Gvozdenovic-Jeremic, J

2001-12-01

An immobilized-cell biofilm reactor was used for the continuous production of lactic acid by Lactobacillus casei subsp. rhamnosus (ATCC 11443). At Iowa State University, a unique plastic composite support (PCS) that stimulates biofilm formation has been developed. The optimized PCS blend for Lactobacillus contains 50% (wt/wt) agricultural products [35% (wt/wt) ground soy hulls, 5% (wt/wt) soy flour, 5% (wt/wt) yeast extract, 5% (wt/wt) dried bovine albumin, and mineral salts] and 50% (wt/wt) polypropylene (PP) produced by high-temperature extrusion. The PCS tubes have a wall thickness of 3.5 mm, outer diameter of 10.5 mm, and were cut into 10-cm lengths. Six PCS tubes, three rows of two parallel tubes, were bound in a grid fashion to the agitator shaft of a 1.2-1 vessel for a New Brunswick Bioflo 3000 fermentor. PCS stimulates biofilm formation, supplies nutrients to attached and suspended cells, and increases lactic acid production. Biofilm thickness on the PCS tubes was controlled by the agitation speed. The PCS biofilm reactor and PP control reactor achieved optimal average production rates of 9.0 and 5.8 g l(-1) h(-1), respectively, at 0.4 h(-1) dilution rate and 125-rpm agitation with yields of approximately 70%.

Characteristics comparison between a cyclotron-based neutron source and KUR-HWNIF for boron neutron capture therapy

NASA Astrophysics Data System (ADS)

Tanaka, H.; Sakurai, Y.; Suzuki, M.; Masunaga, S.; Kinashi, Y.; Kashino, G.; Liu, Y.; Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Maruhashi, A.; Ono, K.

2009-06-01

At Kyoto University Research Reactor Institute (KURRI), 275 clinical trials of boron neutron capture therapy (BNCT) have been performed as of March 2006, and the effectiveness of BNCT has been revealed. In order to further develop BNCT, it is desirable to supply accelerator-based epithermal-neutron sources that can be installed near the hospital. We proposed the method of filtering and moderating fast neutrons, which are emitted from the reaction between a beryllium target and 30-MeV protons accelerated by a cyclotron accelerator, using an optimum moderator system composed of iron, lead, aluminum and calcium fluoride. At present, an epithermal-neutron source is under construction from June 2008. This system consists of a cyclotron accelerator, beam transport system, neutron-yielding target, filter, moderator and irradiation bed. In this article, an overview of this system and the properties of the treatment neutron beam optimized by the MCNPX Monte Carlo neutron transport code are presented. The distribution of biological effect weighted dose in a head phantom compared with that of Kyoto University Research Reactor (KUR) is shown. It is confirmed that for the accelerator, the biological effect weighted dose for a deeply situated tumor in the phantom is 18% larger than that for KUR, when the limit dose of the normal brain is 10 Gy-eq. The therapeutic time of the cyclotron-based neutron sources are nearly one-quarter of that of KUR. The cyclotron-based epithermal-neutron source is a promising alternative to reactor-based neutron sources for treatments by BNCT.

University Research in Support of TREAT Modeling and Simulation, FY 2016

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

DeHart, Mark David

Idaho National Laboratory is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under the Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. In support of this research, INL is working with four universities to explore advanced solution methods that will complement or augment capabilities in MAMMOTH. This report consists of a collection of year end summaries of research from the universities performed inmore » support of TREAT modeling and simulation. This research was led by Prof. Sedat Goluoglu at the University of Florida, Profs. Jim Morel and Jean Ragusa at Texas A&M University, Profs. Benoit Forget and Kord Smith at Massachusetts Institute of Technology, Prof. Leslie Kerby of Idaho State University and Prof. Barry Ganapol of University of Arizona. A significant number of students were supported at various levels though the projects and, for some, also as interns at INL.« less

A Tire Gasification Senior Design Project That Integrates Laboratory Experiments and Computer Simulation

ERIC Educational Resources Information Center

Weiss, Brian; Castaldi, Marco J.

2006-01-01

A reactor to convert waste rubber tires to useful products such as CO and H2, was investigated in a university undergraduate design project. The student worked individually with mentorship from a faculty professor who aided the student with professional critique. The student was able to research the background of the field and conceive of a novel…

Using Biotechnology in the Laboratory: Using an Immobilized-Laccase Reactor-System to Learn about Wastewater Treatment

ERIC Educational Resources Information Center

Genc, Rukan; Rodriguez-Couto, Susana

2009-01-01

This article includes a practical guide, which was used to teach the phenomenon of immobilization of enzymes and their subsequent use for discoloration of dyes to under-graduate students of Biotechnology at the Rovira i Virgili University (Tarragona, Spain). Alginate was selected as a support for the immobilization of laccase. Remazol Brilliant…

Microwave Plasma Based Single-Step Method for Generation of Carbon Nanostructures

DTIC Science & Technology

2013-07-01

Técnico, Technical University of Lisbon, Portugal 2 Mechanical and Aerospace Engeneering , Naval Postgraduate School, Monterey, CA 93943, U.S.A...Plasma environments constitute powerful tools in materials science due to their operation as thermal and chemical reactors. A microwave, atmospheric...applications include electronic devices, transparent conductive films, mechanical devices, chemical sensors, spintronic devices. Moreover, it shows enormous

Lifelong Learning: The Value of an Industrial Internship for a Graduate Student Education

ERIC Educational Resources Information Center

Honda, Gregory S.; Pazmino, Jorge H.; Hickman, Daniel A.; Varma, Arvind

2015-01-01

A chemical engineering PhD student from Purdue University completed an internship at The Dow Chemical Company, evaluating the effect of scale on the hydrodynamics of a trickle bed reactor. A unique aspect of this work was that it arose from an ongoing collaboration, so that the project was within the scope of the graduate student's thesis. This…

To Help Teachers Identify Essential Steps in Planning for Occupational Programs in Home Economics. An Audiotutorial Kit containing Ten Tapes with Reactor Sheets.

ERIC Educational Resources Information Center

Granberg, Grace

The materials for individualized or group instruction feature a workbook type format with correlated tape discussions. The program, developed at a 1967 teacher education workshop at Iowa State University utilizes the expertise of several leaders and resource guest speakers to acquaint preservice and inservice teachers with vocational home…

A Unique Master's Program in Combined Nuclear Technology and Nuclear Chemistry at Chalmers University of Technology, Sweden

NASA Astrophysics Data System (ADS)

Skarnemark, Gunnar; Allard, Stefan; Ekberg, Christian; Nordlund, Anders

2009-08-01

The need for engineers and scientists who can ensure safe and secure use of nuclear energy is large in Sweden and internationally. Chalmers University of Technology is therefore launching a new 2-year master's program in Nuclear Engineering, with start from the autumn of 2009. The program is open to Swedish and foreign students. The program starts with compulsory courses dealing with the basics of nuclear chemistry and physics, radiation protection, nuclear power and reactors, nuclear fuel supply, nuclear waste management and nuclear safety and security. There are also compulsory courses in nuclear industry applications and sustainable energy futures. The subsequent elective courses can be chosen freely but there is also a possibility to choose informal tracks that concentrate on nuclear chemistry or reactor technology and physics. The nuclear chemistry track comprises courses in e.g. chemistry of lanthanides, actinides and transactinides, solvent extraction, radioecology and radioanalytical chemistry and radiopharmaceuticals. The program is finished with a one semester thesis project. This is probably a unique master program in the sense of its combination of deep courses in both nuclear technology and nuclear chemistry.

Thermal storage/discharge performances of Cu-Si alloy for solar thermochemical process

NASA Astrophysics Data System (ADS)

Gokon, Nobuyuki; Yamaguchi, Tomoya; Cho, Hyun-seok; Bellan, Selvan; Hatamachi, Tsuyoshi; Kodama, Tatsuya

2017-06-01

The present authors (Niigata University, Japan) have developed a tubular reactor system using novel "double-walled" reactor/receiver tubes with carbonate molten-salt thermal storage as a phase change material (PCM) for solar reforming of natural gas and with Al-Si alloy thermal storage as a PCM for solar air receiver to produce high-temperature air. For both of the cases, the high heat capacity and large latent heat (heat of solidification) of the PCM phase circumvents the rapid temperature change of the reactor/receiver tubes at high temperatures under variable and uncontinuous characteristics of solar radiation. In this study, we examined cyclic properties of thermal storage/discharge for Cu-Si alloy in air stream in order to evaluate a potentiality of Cu-Si alloy as a PCM thermal storage material. Temperature-increasing performances of Cu-Si alloy are measured during thermal storage (or heat-charge) mode and during cooling (or heat-discharge) mode. A oxidation state of the Cu-Si alloy after the cyclic reaction was evaluated by using electron probe micro analyzer (EPMA).

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

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

Yoon, Dhongik S.; Jo, HangJin; Fu, Wen

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s abilitymore » to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.« less

Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

NASA Astrophysics Data System (ADS)

Muratov, V. G.; Lopatkin, A. V.

An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

Analytic expressions for Atomic Layer Deposition: coverage, throughput, and materials utilization in cross-flow, particle coating, and spatial ALD

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

Yanguas-Gil, Angel; Elam, Jeffrey W.

2014-05-01

In this work, the authors present analytic models for atomic layer deposition (ALD) in three common experimental configurations: cross-flow, particle coating, and spatial ALD. These models, based on the plug-flow and well-mixed approximations, allow us to determine the minimum dose times and materials utilization for all three configurations. A comparison between the three models shows that throughput and precursor utilization can each be expressed by universal equations, in which the particularity of the experimental system is contained in a single parameter related to the residence time of the precursor in the reactor. For the case of cross-flow reactors, the authorsmore » show how simple analytic expressions for the reactor saturation profiles agree well with experimental results. Consequently, the analytic model can be used to extract information about the ALD surface chemistry (e. g., the reaction probability) by comparing the analytic and experimental saturation profiles, providing a useful tool for characterizing new and existing ALD processes. (C) 2014 American Vacuum Society« less

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

DOE PAGES

Yoon, Dhongik S.; Jo, HangJin; Fu, Wen; ...

2017-05-23

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s abilitymore » to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.« less

Neutron interferometry: The pioneering contributions of Samuel A. Werner

NASA Astrophysics Data System (ADS)

Klein, A. G.

2006-11-01

In 1975, Sam Werner, while on the staff of the Scientific Laboratory of the Ford Motor Company, and his collaborators from Purdue University, Roberto Colella and Albert Overhauser, carried out one of the pioneering experiments in neutron interferometry at the 2 MW University of Michigan research reactor. It was the famous COW Experiment [Colella et al., Phys. Rev. Lett. 34 (1975) 1472] on gravitationally induced quantum interference. Shortly thereafter he moved to the University of Missouri in Columbia, to set up a program of neutron scattering research, including neutron interferometry. In the 25 years until his retirement a large number of beautiful experiments have been performed by Sam, with his group, his numerous students and many international collaborators. This work and its history are briefly reviewed in this paper.

High Neutron Fluence Survivability Testing of Advanced Fiber Bragg Grating Sensors

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

Fielder, Robert S.; Klemer, Daniel; Stinson-Bagby, Kelly L.

2004-02-04

The motivation for the reported research was to support NASA space nuclear power initiatives through the development of advanced fiber optic sensors for space-based nuclear power applications. The purpose of the high-neutron fluence testing was to demonstrate the survivability of fiber Bragg grating (FBG) sensors in a fission reactor environment. 520 FBGs were installed in the Ford reactor at the University of Michigan. The reactor was operated for 1012 effective full power hours resulting in a maximum neutron fluence of approximately 5x1019 n/cm2, and a maximum gamma dose of 2x103 MGy gamma. This work is significant in that, to themore » knowledge of the authors, the exposure levels obtained are approximately 1000 times higher than for any previously published experiment. Four different fiber compositions were evaluated. An 87% survival rate was observed for fiber Bragg gratings located at the fuel centerline. Optical Frequency Domain Reflectometry (OFDR), originally developed at the NASA Langley Research Center, can be used to interrogate several thousand low-reflectivity FBG strain and/or temperature sensors along a single optical fiber. A key advantage of the OFDR sensor technology for space nuclear power is the extremely low mass of the sensor, which consists of only a silica fiber 125{mu}m in diameter. The sensors produced using this technology will fill applications in nuclear power for current reactor plants, emerging Generation-IV reactors, and for space nuclear power. The reported research was conducted by Luna Innovations and was funded through a Small Business Innovative Research (SBIR) contract with the NASA Glenn Research Center.« less

Experimentation on the anaerobic filter reactor for biogas production using rural domestic wastewater

NASA Astrophysics Data System (ADS)

Leju Celestino Ladu, John; Lü, Xi-wu; Zhong, Zhaoping

2017-08-01

The biogas production from anaerobic filter (AF) reactor was experimented in Taihu Lake Environmental Engineering Research Center of Southeast University, Wuxi, China. Two rounds of experimental operations were conducted in a laboratory scale at different Hydraulic retention time (HRT) and wastewater temperature. The biogas production rate during the experimentation was in the range of 4.63 to 11.78 L/d. In the first experimentation, the average gas production rate was 10.08 L/d, and in the second experimentation, the average gas production rate was 4.97 L/d. The experimentation observed the favorable Hydraulic Retention Time and wastewater temperature in AF was three days and 30.95°C which produced the gas concentration of 11.78 L/d. The HRT and wastewater temperature affected the efficiency of the AF process on the organic matter removal and nutrients removal as well. It can be deduced from the obtained results that HRT and wastewater temperature directly affects the efficiency of the AF reactor in biogas production. In conclusion, anaerobic filter treatment of organic matter substrates from the rural domestic wastewater increases the efficiency of the AF reactor on biogas production and gives a number of benefits for the management of organic wastes as well as reduction in water pollution. Hence, the operation of the AF reactor in rural domestic wastewater treatment can play an important element for corporate economy of the biogas plant, socio-economic aspects and in the development of effective and feasible concepts for wastewater management, especially for people in rural low-income areas.

Non-Intrusive Velocity Measurements with MTV During DCC Event in the HTTF

NASA Technical Reports Server (NTRS)

Andre, M. A.; Bardet, P. M.; Cadell, S. R.; Woods, B.; Burns, R. A.; Danehy, P. M.

2017-01-01

Velocity profiles are measured using molecular tagging velocimetry (MTV) in the high temperature test facility (HTTF) at Oregon State University during a depressurized conduction cooldown (DCC) event. The HTTF is a quarter scale electrically heated nuclear reactor simulator designed to replicate various accident scenarios. During a DCC, a double ended guillotine break results in the reactor pressure vessel (RPV) depressurizing into the reactor cavity and ultimately leading to air ingress in the reactor core (lock-exchange and gas diffusion). It is critical to understand the resulting buoyancy-driven flow to characterize the reactor self-cooling capacity through natural circulation. During tests conducted at ambient pressure and temperature, the RPV containing helium is opened (via the hot and cold legs) to a large vessel filled with nitrogen to simulate the atmosphere. The velocity profile on the hot leg pipe centerline is recorded at 10 Hz with MTV based on NO tracers. The precision of the velocimetry was measured to be 0.02 m/s in quiescent flow prior to the tests. A helium flow from the RPV is initially observed in the top quarter of the pipe. During the first 20 seconds of the event, helium flows out of the RPV with a maximum velocity below 2 m/s. The velocity profile transitions from parabolic to linear in character and decays slowly over the rest of the recording; peak velocities of 0.2 m/s are observed after 30 min. A counter-flow of nitrogen is also observed intermittently, which occurs at lower velocities (>0.1 m/s).

High Neutron Fluence Survivability Testing of Advanced Fiber Bragg Grating Sensors

NASA Astrophysics Data System (ADS)

Fielder, Robert S.; Klemer, Daniel; Stinson-Bagby, Kelly L.

2004-02-01

The motivation for the reported research was to support NASA space nuclear power initiatives through the development of advanced fiber optic sensors for space-based nuclear power applications. The purpose of the high-neutron fluence testing was to demonstrate the survivability of fiber Bragg grating (FBG) sensors in a fission reactor environment. 520 FBGs were installed in the Ford reactor at the University of Michigan. The reactor was operated for 1012 effective full power hours resulting in a maximum neutron fluence of approximately 5×1019 n/cm2, and a maximum gamma dose of 2×103 MGy gamma. This work is significant in that, to the knowledge of the authors, the exposure levels obtained are approximately 1000 times higher than for any previously published experiment. Four different fiber compositions were evaluated. An 87% survival rate was observed for fiber Bragg gratings located at the fuel centerline. Optical Frequency Domain Reflectometry (OFDR), originally developed at the NASA Langley Research Center, can be used to interrogate several thousand low-reflectivity FBG strain and/or temperature sensors along a single optical fiber. A key advantage of the OFDR sensor technology for space nuclear power is the extremely low mass of the sensor, which consists of only a silica fiber 125μm in diameter. The sensors produced using this technology will fill applications in nuclear power for current reactor plants, emerging Generation-IV reactors, and for space nuclear power. The reported research was conducted by Luna Innovations and was funded through a Small Business Innovative Research (SBIR) contract with the NASA Glenn Research Center.

Estimation of the ionic charge of non-metallic species into an electrical discharge through a web application

NASA Astrophysics Data System (ADS)

Pérez Gutiérrez, B. R.; Vera-Rivera, F. H.; Niño, E. D. V.

2016-08-01

Estimate the ionic charge generated in electrical discharges will allow us to know more accurately the concentration of ions implanted on the surfaces of nonmetallic solids. For this reason, in this research a web application was developed to allow us to calculate the ionic charge generated in an electrical discharge from the experimental parameters established in an ion implantation process performed in the JUPITER (Joint Universal Plasma and Ion Technologies Experimental Reactor) reactor. The estimated value of the ionic charge will be determined from data acquired on an oscilloscope, during startup and shutdown of electrical discharge, which will then be analyzed and processed. The study will provide best developments with regard to the application of ion implantation in various industrial sectors.

The elemental move characteristic of nickel-based alloy in molten salt corrosion by using nuclear microprobe

NASA Astrophysics Data System (ADS)

Lei, Qiantao; Liu, Ke; Gao, Jie; Li, Xiaolin; Shen, Hao; Li, Yan

2017-08-01

Nickel-based alloys as candidate materials for Thorium Molten Salt Reactor (TMSR), need to be used under high temperature in molten salt environment. In order to ensure the safety of the reactor running, it is necessary to study the elemental move characteristic of nickel-based alloys in the high temperature molten salts. In this work, the scanning nuclear microprobe at Fudan University was applied to study the elemental move. The Nickel-based alloy samples were corroded by molten salt at different temperatures. The element concentrations in the Nickel-based alloys samples were determined by the scanning nuclear microprobe. Micro-PIXE results showed that the element concentrations changed from the interior to the exterior of the alloy samples after the corrosion.

Measuring Theta_13 at Daya Bay

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

Lau, Kwong

2014-03-14

We measured the neutrino mixing angle, theta13, presumably related to the preponderance of matter over antimatter in our universe with high precision. We determined theta13 by measuring the disappearance of neutrinos from a group of six nuclear reactors. The target, located inside a mountain at about 2 km from the reactors, is 80 tons of liquid scintillator doped with trace amount of Gadolinium to increase its neutron detection efficiency. The neutrino flux is measured by the inverse beta-decay reaction where the final-state particles are detected by the liquid scintillator. The measured value of theta13, based on data collected over 3more » years, is large, around 8 degrees, rendering the measurement of the parameter related to matter-antimatter asymmetry in future long baseline neutrino experiments easier.« less

Test of a prototype neutron spectrometer based on diamond detectors in a fast reactor

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

Osipenko, M.; Ripani, M.; Ricco, G.

2015-07-01

A prototype of neutron spectrometer based on diamond detectors has been developed. This prototype consists of a {sup 6}Li neutron converter sandwiched between two CVD diamond crystals. The radiation hardness of the diamond crystals makes it suitable for applications in low power research reactors, while a low sensitivity to gamma rays and low leakage current of the detector permit to reach good energy resolution. A fast coincidence between two crystals is used to reject background. The detector was read out using two different electronic chains connected to it by a few meters of cable. The first chain was based onmore » conventional charge-sensitive amplifiers, the other used a custom fast charge amplifier developed for this purpose. The prototype has been tested at various neutron sources and showed its practicability. In particular, the detector was calibrated in a TRIGA thermal reactor (LENA laboratory, University of Pavia) with neutron fluxes of 10{sup 8} n/cm{sup 2}s and at the 3 MeV D-D monochromatic neutron source named FNG (ENEA, Rome) with neutron fluxes of 10{sup 6} n/cm{sup 2}s. The neutron spectrum measurement was performed at the TAPIRO fast research reactor (ENEA, Casaccia) with fluxes of 10{sup 9} n/cm{sup 2}s. The obtained spectra were compared to Monte Carlo simulations, modeling detector response with MCNP and Geant4. (authors)« less

Students' Pilot Laboratory for Homogeneous Chemical Reactor Analysis and Design in Olive Mill Wastewater Treatment

ERIC Educational Resources Information Center

Ochando-Pulido, J. M.

2017-01-01

The Chemical Engineering Department at the University of Granada have endeavored to make a number of high quality experiments to familiarize our students with our latest research and also scale-up of processes. A pilot-scale wastewater treatment plant was set-up to give students a close practical view of the treatments of effluents by-produced in…

Bulk shielding facility quarterly report, October, November, and December 1976

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

Hurt, III, S. S.; Lance, E. D.; Thomas, J. R.

1977-08-01

The BSR operated at an average power level of 1,836 kw for 78.01 percent of the time during October, November, and December. Water-quality control in both the reactor primary and secondary cooling systems was satisfactory. The PCA was used in training programs and was operated on two occasions when the University of Kentucky students actively participated in training laboratories.

Decommissioning of German Nuclear Research Facilities under the Governance of the Federal Ministry of Education and Research

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

Weigl, M.

2008-07-01

Since the announcement of the first nuclear program in 1956, nuclear R and D in Germany has been supported by the Federal Government under four nuclear programs and later on under more general energy R and D programs. The original goal was to help German industry to achieve safe, low-cost generation of energy and self-sufficiency in the various branches of nuclear technology, including the fast breeder reactor and the fuel cycle. Several national research centers were established to host or operate experimental and demonstration plants. These are mainly located at the sites of the national research centers at Juelich andmore » Karlsruhe. In the meantime, all these facilities were shut down and most of them are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactor with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. For two other projects the return to 'green field' sites will be reached by the end of this decade. These are the dismantling of the Multi-Purpose Research Reactor (MZFR) and the Compact Sodium Cooled Reactor (KNK) both located at the Forschungszentrum Karlsruhe. Within these projects a lot of new solutions und innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). For example, high performance underwater cutting technologies like plasma arc cutting and contact arc metal cutting. (authors)« less

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

Adaptive Neural Network Algorithm for Power Control in Nuclear Power Plants

NASA Astrophysics Data System (ADS)

Masri Husam Fayiz, Al

2017-01-01

The aim of this paper is to design, test and evaluate a prototype of an adaptive neural network algorithm for the power controlling system of a nuclear power plant. The task of power control in nuclear reactors is one of the fundamental tasks in this field. Therefore, researches are constantly conducted to ameliorate the power reactor control process. Currently, in the Department of Automation in the National Research Nuclear University (NRNU) MEPhI, numerous studies are utilizing various methodologies of artificial intelligence (expert systems, neural networks, fuzzy systems and genetic algorithms) to enhance the performance, safety, efficiency and reliability of nuclear power plants. In particular, a study of an adaptive artificial intelligent power regulator in the control systems of nuclear power reactors is being undertaken to enhance performance and to minimize the output error of the Automatic Power Controller (APC) on the grounds of a multifunctional computer analyzer (simulator) of the Water-Water Energetic Reactor known as Vodo-Vodyanoi Energetichesky Reaktor (VVER) in Russian. In this paper, a block diagram of an adaptive reactor power controller was built on the basis of an intelligent control algorithm. When implementing intelligent neural network principles, it is possible to improve the quality and dynamic of any control system in accordance with the principles of adaptive control. It is common knowledge that an adaptive control system permits adjusting the controller’s parameters according to the transitions in the characteristics of the control object or external disturbances. In this project, it is demonstrated that the propitious options for an automatic power controller in nuclear power plants is a control system constructed on intelligent neural network algorithms.

Operation and reactivity measurements of an accelerator driven subcritical TRIGA reactor

NASA Astrophysics Data System (ADS)

O'Kelly, David Sean

Experiments were performed at the Nuclear Engineering Teaching Laboratory (NETL) in 2005 and 2006 in which a 20 MeV linear electron accelerator operating as a photoneutron source was coupled to the TRIGA (Training, Research, Isotope production, General Atomics) Mark II research reactor at the University of Texas at Austin (UT) to simulate the operation and characteristics of a full-scale accelerator driven subcritical system (ADSS). The experimental program provided a relatively low-cost substitute for the higher power and complexity of internationally proposed systems utilizing proton accelerators and spallation neutron sources for an advanced ADSS that may be used for the burning of high-level radioactive waste. Various instrumentation methods that permitted ADSS neutron flux monitoring in high gamma radiation fields were successfully explored and the data was used to evaluate the Stochastic Pulsed Feynman method for reactivity monitoring.

Status of reduced enrichment programs for research reactors in Japan

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

Kanda, Keiji; Nishihara, Hedeaki; Shirai, Eiji

1997-08-01

The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for themore » full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.« less

Assessment of the National Research Universal Reactor Proposed New Stack Sampling Probe Location for Compliance with ANSI/HPS N13.1-1999

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

Glissmeyer, John A.; Antonio, Ernest J.; Flaherty, Julia E.

2016-02-29

This document reports on a series of tests conducted to assess the proposed air sampling location for the National Research Universal reactor (NRU) complex exhaust stack, located in Chalk River, Ontario, Canada, with respect to the applicable criteria regarding the placement of an air sampling probe. Due to the age of the equipment in the existing monitoring system, and the increasing difficulty in acquiring replacement parts to maintain this equipment, a more up-to-date system is planned to replace the current effluent monitoring system, and a new monitoring location has been proposed. The new sampling probe should be located within themore » exhaust stack according to the criteria established by the American National Standards Institute/Health Physics Society (ANSI/HPS) N13.1-1999, Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stack and Ducts of Nuclear Facilities. These criteria address the capability of the sampling probe to extract a sample that represents the effluent stream. The internal Pacific Northwest National Laboratory (PNNL) project for this task was 65167, Atomic Energy Canada Ltd. Chalk River Effluent Duct Flow Qualification. The testing described in this document was guided by the Test Plan: Testing of the NRU Stack Air Sampling Position (TP-STMON-032).« less

Ion source development for a photoneutralization based NBI system for fusion reactors

NASA Astrophysics Data System (ADS)

Simonin, A.; de Esch, H. P. L.; Garibaldi, P.; Grand, C.; Bechu, S.; Bès, A.; Lacoste, A.

2015-04-01

The next step after ITER is to demonstrate the viability and generation of electricity by a future fusion reactor (DEMO). The specifications required to operate an NBI system on DEMO are very demanding. The system has to provide a very high level of power and energy, ~100MW of D° beam at 1MeV, including high wall-plug efficiency (η > 60%). For this purpose, a new injector concept, called Siphore, is under investigation between CEA and French universities. Siphore is based on the stripping of the accelerated negative ions by photo-detachment provided by several Fabry-Perot cavities (3.5MW of light power per cavity) implemented along the D- beam. The beamline is designed to be tall and narrow in order that the photon flux overlaps the entire negative ion beam. The paper will describe the present R&D at CEA which addresses the development of an ion source and pre-accelerator prototypes for Siphore, the main goal being to produce an intense negative ion beam sheet. The negative ion source Cybele is based on a magnetized plasma column where hot electrons are emitted from the source center. Parametric studies of the source are performed using Langmuir probes in order to characterize the plasma and to compare with numerical models being developed in French universities.

Utility operations review of North Carolina State University BSNE curriculum

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

Bishop, E.A.; Faggart, E.M.; Jackson, G.D.

1988-01-01

The industry advisors group of the North Carolina State University (NCSU) Department of Nuclear Engineering raised the question of how well the curriculum for a bachelor of science in nuclear engineering (BSNE) meets the needs of educating students to enter the nuclear operations field. The concern was that the nuclear industry has evolved from a design to an operations mode, but that the BSNE curriculum may not have responded to this evolution. To address this issue, a group of four persons qualified as senior reactor operators with operational experience from different utilities was selected. The authors are the members ofmore » this review group. All are degreed personnel, with three BSNE graduates from NCSU, and all have participated in nuclear plant startups and currently work at nuclear plant sites. The group prepared by reviewing the curriculum before arriving on campus, including the report developed for the Accreditation Board for Engineering and Technology. During our two-day campus visit, we reviewed course materials, interviewed professors, and toured laboratory and reactor facilities in order to get more insight into the breadth and thrust of the BSNE curriculum. The observations and recommendations contained in this paper were developed based on these reviews and discussions and represent the opinions of the authors and not necessarily their companies.« less

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 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

The Japanese aerial attack on Hanford Engineer Works

NASA Astrophysics Data System (ADS)

Clark, Charles W.

The day before the Pearl Harbor attack, December 6, 1941, the University of Chicago Metallurgical Laboratory was given four goals: design a plutonium (Pu) bomb; produce Pu by irradiation of uranium (U); extract Pu from the irradiated U; complete this in time to be militarily significant. A year later the first controlled nuclear chain reaction was attained in Chicago Pile 1 (CP-1). In January 1943, Hanford, WA was chosen as the site of the Pu factory. Neutron irradiation of 238U was to be used to make 239Pu. This was done by a larger version of CP-1, Hanford Reactor B, which went critical in September 1944. By July 1945 it had made enough Pu for two bombs: one used at the Trinity test in July; the other at Nagasaki, Japan in August. I focus on an ironic sidelight to this story: disruption of hydroelectric power to Reactor B by a Japanese fire balloon attack on March 10, 1945. This activated the costly coal-fired emergency backup plant to keep the reactor coolant water flowing, thwarting disaster and vindicating the conservative design of Hanford Engineer Works. Management of the Hanford Engineer Works in World War II, H. Thayer (ASCE Press 1996).

Trace-element characterization of evidential cannabis sative samples using k{sub 0}-standardization methodology

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

Henderson, D.P. Jr.; Vernetson, W.G.; Ratner, R.T.

The University of Florida Training Reactor (UFTR) facilities including the analytical laboratory are used for a wide range of educational, research, training, and service functions. The UFTR is a 100-kW light-water-cooled, graphite-and-water-moderated modified Argonaut-type reactor. The UFTR utilizes high enriched plate-type fuel in a two-slab arrangement and operates at a 100-kW power level. Since first licensed to operate at 10 kW in 1959, this nonpower reactor facility has had an active but evolving record of continuous service to a wide range of academic, utility, and community users. The services of the UFTR have also been used by various state authoritiesmore » in criminal investigations. Because of its relatively low power and careful laboratory analyses, the UFTR neutron flux characteristics in several ports are not only well characterized but they are also quite invariant with time. As a result, such a facility is well-suited to the application of the multielement analysis using the k{sub o}-standardization method of neutron activation analysis. The analysis of untreated evidential botanical samples presented a unique opportunity to demonstrate implementation of this method at the UFTR facilities.« less

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

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

Jacobsen, George; Back, Christina

2015-10-30

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

Israeli co-retorting of coal and oil shale would break even at 22/barrel

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

Not Available

Work is being carried out at the Hebrew University of Jerusalem on co-retorting of coal and oil shale. The work is funded under a cooperative agreement with the US Department of Energy. The project is exploring the conversion of US eastern high-sulfur bituminous coal in a split-stage, fluidized-bed reactor. Pyrolysis occurs in the first stage and char combustion in the second stage. These data for coal will be compared with similar data from the same reactor fueled by high-sulfur eastern US oil shale and Israeli oil shales. The project includes research at three major levels: pyrolysis in lab-scale fluidized-bed reactor;more » retorting in split-stage, fluidized-bed bench-scale process (1/4 tpd); and scale-up, preparation of full-size flowchart, and economic evaluation. In the past year's research, a preliminary economic evaluation was completed for a scaled-up process using a feed of high-sulfur coal and carbonate-containing Israeli oil shale. A full-scale plant in Israel was estimated to break even at an equivalent crude oil price of $150/ton ($22/barrel).« less

Novel, Integrated Reactor/Power Conversion System (LMR-AMTEC)

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

Dmitry V. Paramonov, Lead Collaborator

2001-07-31

The overall objective of NERI Project Number 99-0198 is to assess the technical and economic feasibility, develop engineering solutions and determine a range of potential applications for ''Novel Integrated Reactor/Energy conversion Systems''. The near term goal is the design of a power supply for developing countries in remote locations in a proliferation resistant, reliable and economical way. The heart of the concept is the use of a single loop liquid metal fast reactor (LMR) with conversion of the heat directly into electricity in a Alkali Metal Thermal to Electric Converter (AMTEC). The first year of the project focused on themore » feasibility issues with a long life, high temperature liquid metal-cooled core; selection of the working fluid, core-to-AMTEC coupling scheme and interface parameters; and, energy conversion systems design and performance. Report Number STD-ES-01-0028, Revision 0, dated July 31, 2001, summarizes the work performed by Westinghouse personnel in Year One and report number UNM-ISNPS-3-2000, dated October 2000, summarizes the work performed by the Institute for Space and Nuclear Power Studies at the University of New Mexico in Year One.« less

Epithermal neutron beam for BNCT research at Washington State University

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

Venhuizen, J.R.; Nigg, D.W.; Wheeler, F.J.

1999-09-01

Veterinary radiation oncology researchers at the Washington State University (WSU) School of Veterinary Medicine have made major contributions to the understanding of the in-vivo radiobiology of boron neutron capture therapy (BNCT) over the years. Recent attention has been focused on the development of a more convenient and cost-effective local epithermal-neutron beam facility for BNCT research and boronated pharmaceutical screening in large-animal models at WSU. The design of such a facility, to be installed in the thermal column region of the TRIGA research reactor at WSU, was performed in a collaborative effort of SWU and the Idaho National Engineering and Environmentalmore » Laboratory. Construction is now underway.« less

Vanderbilt University Gamma Irradiation of Nano-modified Concrete (2017 Milestone Report)

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

Deichert, Geoffrey G.; Linton, Kory D.; Terrani, Kurt A.

This document outlines the irradiation of concrete specimens in the Gamma Irradiation Facility in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Two gamma irradiation runs were performed in July of 2017 on 18 reference mortar bar specimens, 26 reference cement paste bar specimens, and 28 reference cement paste tab specimens to determine the dose and temperature response of the specimens in the gamma irradiation environment. Specimens from the first two gamma irradiations were surveyed and released to Vanderbilt University. The temperature and dose information obtained informs the test parameters of the final two gamma irradiationsmore » of nano-modified concrete planned for FY 2018.« less

PREFACE: SANS-YuMO User Meeting at the Start-up of Scientific Experiments on the IBR-2M Reactor: Devoted to the 75th anniversary of Yu M Ostanevich's birth

NASA Astrophysics Data System (ADS)

Gordely, Valentin; Kuklin, Alexander; Balasoiu, Maria

2012-03-01

The Second International Workshop 'SANS-YuMO User Meeting at the Start-up of Scientific Experiments on the IBR-2M Reactor', devoted to the 75th anniversary of the birth of Professor Yu M Ostanevich (1936-1992), an outstanding neutron physicist and the founder of small-angle neutron scattering (field, group, and instrument) at JINR FLNPh, was held on 27-30 May at the Frank Laboratory of Neutron Physics. The first Workshop was held in October 2006. Research groups from different neutron centers, universities and research institutes across Europe presented more than 35 oral and poster presentations describing scientific and methodological results. Most of them were obtained with the help of the YuMO instrument before the IBR-2 shutdown in 2006. For the last four years the IBR-2 reactor has been shut down for refurbishment. At the end of 2010 the physical launch of the IBR-2M reactor was finally realized. Nowadays the small-angle neutron scattering (SANS) technique is applied to a wide range of scientific problems in condensed matter, soft condensed matter, biology and nanotechnology, and despite the fact that there are currently over 30 SANS instruments in operation worldwide at both reactor and spallation sources, the demand for beam-time is considerably higher than the time available. It must be remembered, however, that as the first SANS machine on a steady-state reactor was constructed at the Institute Laue Langevin, Grenoble, the first SANS instrument on a 'white' neutron pulsed beam was accomplished at the Joint Institute for Nuclear Research at the IBR-30 reactor, beamline N5. During the meeting Yu M Ostanevich's determinative and crucial contribution to the construction of spectrometers at the IBR-2 high-pulsed reactor was presented, as well as his contribution to the development of the time-of-flight (TOF) small-angle scattering technique, and a selection of other scientific areas. His leadership and outstanding scientific achievements in applications of the Mossbauer effect in physics and chemistry, in SANS studies of polyelectrolytes, small molecules, fractals, metallic glasses, macromolecules, polymers, etc., were recognized by a number of awards including the State Prize of the Russian Federation in 2000. The scientific program of the workshop focused on fundamental and methodical research at the YuMO spectrometer and developments of the SANS instrument at the modernized IBR-2M reactor. We recall that the acronym YuMO of the small-angle neutron scattering spectrometer (MURN), was given in honor of Yu M Ostanevich. One of the most important objectives of this user meeting was to discuss the further development possibilities of the YuMO spectrometer with experts, in the frame of a SANS YuMO Round Table, taking into account the specific performance of the modernized YuMO SANS instrument, and the scientific and technical requests of the instrument's users. Highlights on modern achievements in nanoscience, polymers and biology were other significant goals of the meeting. The plenary invited talks were presented by leading scientists in small-angle neutron scattering and soft condensed matter, including members of the Russian Academy of Sciences: Prof. Heinrich Stuhrmann, Prof. Alexei Khokhlov, Prof. Jose Teixeira, Prof. Alexander Ozerin, Prof. Albrecht Wiedenmann, etc. There were 27 oral talks given and 32 posters presented by 92 participants from 12 countries: Czech Republic, Egypt, France, Germany, Hungary, Moldova, Mongolia, Poland, Romania, Russian Federation, Slovak Republic, and Ukraine. The workshop was organized with the financial support of the Frank Laboratory of Neutron Physics (Joint Institute for Nuclear Research), Horia Hulubei National Institute of Physics and Nuclear Engineering - IFIN HH (Romania), Institute of Macromolecular Chemistry AS CR (Czech Republic), and Comenius University (Slovakia). V Gordeliy, A Kuklin and M Balasoiu SANSgroup Participants of the meeting The PDF also contains additional photographs from the meeting.

Decontamination and decommissioning of the Mayaguez (Puerto Rico) facility

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

Jackson, P.K.; Freemerman, R.L.

1989-11-01

On February 6, 1987 the US Department of Energy (DOE) awarded the final phase of the decontamination and decommissioning of the nuclear and reactor facilities at the Center for Energy and Environmental Research (CEER), in Mayaguez, Puerto Rico. Bechtel National, Inc., was made the decontamination and decommissioning (D and D) contractor. The goal of the project was to enable DOE to proceed with release of the CEER facility for use by the University of Puerto Rico, who was the operator. This presentation describes that project and lesson learned during its progress. The CEER facility was established in 1957 as themore » Puerto Rico Nuclear Center, a part of the Atoms for Peace Program. It was a nuclear training and research institution with emphasis on the needs of Latin America. It originally consisted of a 1-megawatt Materials Testing Reactor (MTR), support facilities and research laboratories. After eleven years of operation the MTR was shutdown and defueled. A 2-megawatt TRIGA reactor was installed in 1972 and operated until 1976, when it woo was shutdown. Other radioactive facilities at the center included a 10-watt homogeneous L-77 training reactor, a natural uranium graphite-moderated subcritical assembly, a 200KV particle accelerator, and a 15,000 Ci Co-60 irradiation facility. Support facilities included radiochemistry laboratories, counting rooms and two hot cells. As the emphasis shifted to non-nuclear energy technology a name change resulted in the CEER designation, and plans were started for the decontamination and decommissioning effort.« less

The extraction of bitumen from western oil sands: Volume 2. Final report

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

Oblad, A.G.; Dahlstrom, D.A.; Deo, M.D.

1997-11-26

The program is composed of 20 projects, of which 17 are laboratory bench or laboratory pilot scale processes or computer process simulations that are performed in existing facilities on the University of Utah campus in north-east Salt Lake City. These tasks are: (1) coupled fluidized-bed bitumen recovery and coked sand combustion; (2) water-based recovery of bitumen; (3) oil sand pyrolysis in a continuous rotary kiln reactor; (4) oil sand pyrolysis in a large diameter fluidized bed reactor; (5) oil sand pyrolysis in a small diameter fluidized bed reactor; (6) combustion of spent sand in a transport reactor; (7) recovery andmore » upgrading of oil sand bitumen using solvent extraction methods; (8) fixed-bed hydrotreating of Uinta Basin bitumens and bitumen-derived hydrocarbon liquids; (9) ebullieted bed hydrotreating of bitumen and bitumen derived liquids; (10) bitumen upgrading by hydropyrolysis; (11) evaluation of Utah`s major oil sand deposits for the production of asphalt, high-energy jet fuels and other specialty products; (12) characterization of the bitumens and reservoir rocks from the Uinta Basin oil sand deposits; (13) bitumen upgrading pilot plant recommendations; (14) liquid-solid separation and fine tailings thickening; (15) in-situ production of heavy oil from Uinta Basin oil sand deposits; (16) oil sand research and development group analytical facility; and (17) process economics. This volume contains reports on nine of these projects, references, and a bibliography. 351 refs., 192 figs., 65 tabs.« less

The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

DOE PAGES

Garrison, L. M.; Zenobia, Samuel J.; Egle, Brian J.; ...

2016-08-01

The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000°C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ionmore » gun can irradiate the samples with ion currents of 20 μA–500 μA; the typical current used is 72 μA, which is an average flux of 9 × 10 14 ions/(cm 2 s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. In conclusion, the MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.« less

The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions.

PubMed

Garrison, L M; Zenobia, S J; Egle, B J; Kulcinski, G L; Santarius, J F

2016-08-01

The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ion gun can irradiate the samples with ion currents of 20 μA-500 μA; the typical current used is 72 μA, which is an average flux of 9 × 10(14) ions/(cm(2) s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.

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

The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

NASA Astrophysics Data System (ADS)

Garrison, L. M.; Zenobia, S. J.; Egle, B. J.; Kulcinski, G. L.; Santarius, J. F.

2016-08-01

The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ion gun can irradiate the samples with ion currents of 20 μA-500 μA; the typical current used is 72 μA, which is an average flux of 9 × 1014 ions/(cm2 s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.

Nuclear Data Needs for Generation IV Nuclear Energy Systems

NASA Astrophysics Data System (ADS)

Rullhusen, Peter

2006-04-01

Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S. Kozier, G. R. Dyck. The lead cooled fast reactor benchmark BREST-300: analysis with sensitivity method / V. Smirnov ... [et al.]. Sensitivity analysis of neutron cross-sections considered for design and safety studies of LFR and SFR generation IV systems / K. Tucek, J. Carlsson, H. Wider -- Experiments. INL capabilities for nuclear data measurements using the Argonne intense pulsed neutron source facility / J. D. Cole ... [et al.]. Cross-section measurements in the fast neutron energy range / A. Plompen. Recent measurements of neutron capture cross sections for minor actinides by a JNC and Kyoto University Group / H. Harada ... [et al.]. Determination of minor actinides fission cross sections by means of transfer reactions / M. Aiche ... [et al.] -- Evaluated data libraries. Nuclear data services from the NEA / H. Henriksson, Y. Rugama. Nuclear databases for energy applications: an IAEA perspective / R. Capote Noy, A. L. Nichols, A. Trkov. Nuclear data evaluation for generation IV / G. Noguère ... [et al.]. Improved evaluations of neutron-induced reactions on americium isotopes / P. Talou ... [et al.]. Using improved ENDF-based nuclear data for candu reactor calculations / J. Prodea. A comparative study on the graphite-moderated reactors using different evaluated nuclear data / Do Heon Kim ... [et al.].

Model validation using CFD-grade experimental database for NGNP Reactor Cavity Cooling Systems with water and air

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

Manera, Annalisa; Corradini, Michael; Petrov, Victor

This project has been focused on the experimental and numerical investigations of the water-cooled and air-cooled Reactor Cavity Cooling System (RCCS) designs. At this aim, we have leveraged an existing experimental facility at the University of Wisconsin-Madison (UW), and we have designed and built a separate effect test facility at the University of Michigan. The experimental facility at UW has underwent several upgrades, including the installation of advanced instrumentation (i.e. wire-mesh sensors) built at the University of Michigan. These provides highresolution time-resolved measurements of the void-fraction distribution in the risers of the water-cooled RCCS facility. A phenomenological model has beenmore » developed to assess the water cooled RCCS system stability and determine the root cause behind the oscillatory behavior that occurs under normal two-phase operation. Testing under various perturbations to the water-cooled RCCS facility have resulted in changes in the stability of the integral system. In particular, the effects on stability of inlet orifices, water tank volume have and system pressure been investigated. MELCOR was used as a predictive tool when performing inlet orificing tests and was able to capture the Density Wave Oscillations (DWOs) that occurred upon reaching saturation in the risers. The experimental and numerical results have then been used to provide RCCS design recommendations. The experimental facility built at the University of Michigan was aimed at the investigation of mixing in the upper plenum of the air-cooled RCCS design. The facility has been equipped with state-of-theart high-resolution instrumentation to achieve so-called CFD grade experiments, that can be used for the validation of Computational Fluid Dynanmics (CFD) models, both RANS (Reynold-Averaged) and LES (Large Eddy Simulations). The effect of risers penetration in the upper plenum has been investigated as well.« less

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

Stewart Zweben; Samuel Cohen; Hantao Ji

Small ''concept exploration'' experiments have for many years been an important part of the fusion research program at the Princeton Plasma Physics Laboratory (PPPL). this paper describes some of the present and planned fusion concept exploration experiments at PPPL. These experiments are a University-scale research level, in contrast with the larger fusion devices at PPPL such as the National Spherical Torus Experiment (NSTX) and the Tokamak Fusion Test Reactor (TFTR), which are at ''proof-of-principle'' and ''proof-of-performance'' levels, respectively.

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

Pomorski, Michal; Mer-Calfati, Christine; Foulon, Francois

Diamond exhibits a combination of properties which makes it attractive for neutron detection in hostile conditions. In the particular case of detection in a nuclear reactor, it is resilient to radiation, exhibits a natural low sensitivity to gamma rays, and its small size (as compared with that of gas ionisation chambers) enables fluency monitoring with a high position resolution. We report here on the use of synthetic CVD diamond as a solid state micro-fission chamber with U-235 converting material for in-core thermal neutron monitoring. Two types of thin diamond detectors were developed for this application. The first type of detectormore » is fabricated using thin diamond membrane obtained by etching low-cost commercially available single crystal CVD intrinsic diamond, so called 'optical grade' material. Starting from a few hundred of micrometre thick samples, the sample is sliced with a laser and then plasma etched down to a few tenths of micrometre. Here we report the result obtained with a 17 μm thick device. The detection surface of this detector is equal to 1 mm{sup 2}. Detectors with surfaces up to 1 cm{sup 2} can be fabricated with this technique. The second type of detector is fabricated by growing successively two thin films of diamond, by the microwave enhanced chemical vapour deposition technique, on HPHT single crystal diamond. A first, a film of boron doped (p+) single crystal diamond, a few microns thick, is deposited. Then a second film of intrinsic diamond with a thickness of a few tens of microns is deposited. This results in a P doped, Intrinsic, Metal structure (PIM) structure in which the intrinsic volume id the active part of the detector. Here we report the results obtained with a 20 μm thick intrinsic whose detection surface is equal to 0.5 mm{sup 2}, with the possibility to enlarge the surface of the detector up to 1 cm{sup 2}. These two types of detector were tested at the VR-1 research reactor at the Czech Technical University in Prague. The Training Reactor VR-1 is a pool type (light water) reactor based on UO{sub 2} low enriched uranium. It has a nominal power of 1 kW, and can be operated for a short period up to 5 kW. The arrangement of the reactor pool reactor facilitates access to the core, setting and removal of various experimental samples and detectors, and safe and easy handling of fuel assemblies. The reactor is equipped with two horizontal channels (radial and tangential) and 10 vertical channels, of varying diameters, which can be loaded into various core positions, and one pneumatic transfer system. It is also equipped with several specifically designed educational instrumentation systems that can be used to supply complementary measurements and characterization around the reactor. The reactor is operated by the Department of Nuclear Reactors of the Faculty of Nuclear Sciences and Physical Engineering of the Czech Technical University in Prague. The two detectors were placed in-core through one of the vertical insertion channel. They were coupled to remote placed (5 m BNC cable) classical nuclear charge sensitive electronics. Detection properties of both sensors, including: pulse height spectra of U-235 fission fragments (response linearity with neutron flux, count rate, gamma background, were evaluated varying the power of the reactor from 0.005 W to 500 W. The evolution of the counting rate of the thinned optical grade detector as a function of counting rate of a gas ionization chamber used currently for reactor monitoring shows the very good linearity of the detector over the 5 decades. Similar results were obtained with the PIM detector. Additionally fast transient current signals of the detectors were recorded on a digital storage oscilloscope (DSO) using broad-band amplifier and with a simple bias-T, showing potential use of such sensors for neutron counting with no need of an amplification stage, since non-amplified signals from fission fragments exceeded 4 mV in amplitude. Therefore, one can think of simple neutron counting system by feeding diamond detectors signals directly to the low threshold discriminators. The results obtained on the VR1 will be described and discussed in detail in the paper and associated presentation. The results demonstrate that diamond micro-fission chambers can be used for in-core neutron monitoring, where robust, simple and compact devices are required.« less

Nano-Disperse Borides and Carbides: Plasma Technology Production, Specific Properties, Economic Evaluation

NASA Astrophysics Data System (ADS)

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

2016-04-01

The experience of production and study on properties of nano-disperse chromium and titanium borides and carbides, and silicon carbide has been generalized. The structure and special service aspects of utilized plasma-metallurgical complex equipped with a three-jet direct-flow reactor with a capacity of 150 kW have been outlined. Processing, heat engineering and service life characteristics of the reactor are specified. The synthesis parameters of borides and carbides, as well as their basic characteristics in nano-disperse condition and their production flow diagram are outlined. Engineering and economic performance of synthesizing borides in laboratory and industrial conditions is assessed, and the respective segment of the international market as well. The work is performed at State Siberian Industrial University as a project part of the State Order of Ministry of Science and Education of the Russian Federation No. 11.1531/2014/K.

Measurement of airborne 131I, 134)Cs and 137Cs due to the Fukushima reactor incident in Milan (Italy).

PubMed

Clemenza, M; Fiorini, E; Previtali, E; Sala, E

2012-12-01

After the earthquake and the tsunami occurred in Japan on March 2011, four of the Fukushima reactors had released in air a large amount of radioactive isotopes that diffused all over the world. The presence of airborne (131)I, (134)Cs, and (137)Cs in air particulate due to this accident were detected and measured in the Low Radioactivity Laboratory operating in the Department of Environmental Sciences of the University of Milano-Bicocca. The sensitivity of the detecting apparatus is of 0.2 uBq/m(3) of air. Concentration and time distribution of these radiocontaminations ranging from a few to 400 uBq/m(3) for the (131)I and of a few tens of uBq/m(3) for the (137)Cs and (134)Cs. Copyright © 2011 Elsevier Ltd. All rights reserved.

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

Bily, T.

Thermoluminescent dosimeters represent very useful tool for gamma fields parameters measurements at nuclear research reactors, especially at zero power ones. {sup 7}LiF:Mg,Ti and {sup 7}LiF:Mg,Cu,P type TL dosimeters enable determination of only gamma component in mixed neutron - gamma field. At VR-1 reactor operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague the integral characteristics of gamma rays field were investigated, especially its spatial distribution and time behaviour, i.e. the non-saturated delayed gamma ray emission influence. Measured spatial distributions were compared with monte carlo code MCNP5 calculations. Although MCNP cannot generate delayedmore » gamma rays from fission, the relative gamma dose rate distribution is within {+-} 15% with measured values. The experiments were carried out with core configuration C1 consisting of LEU fuel IRT-4M (19.7 %). (author)« less

MCNP study for epithermal neutron irradiation of an isolated liver at the Finnish BNCT facility.

PubMed

Kotiluoto, P; Auterinen, I

2004-11-01

A successful boron neutron capture treatment (BNCT) of a patient with multiple liver metastases has been first given in Italy, by placing the removed organ into the thermal neutron column of the Triga research reactor of the University of Pavia. In Finland, FiR 1 Triga reactor with an epithermal neutron beam well suited for BNCT has been extensively used to irradiate patients with brain tumors such as glioblastoma and recently also head and neck tumors. In this work we have studied by MCNP Monte Carlo simulations, whether it would be beneficial to treat an isolated liver with epithermal neutrons instead of thermal ones. The results show, that the epithermal field penetrates deeper into the liver and creates a build-up distribution of the boron dose. Our results strongly encourage further studying of irradiation arrangement of an isolated liver with epithermal neutron fields.

Fatigue behavior of type 316 stainless steel following neutron irradiation inducing helium

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

Grossbeck, M.L.; Liu, K.C.

1980-01-01

Since a tokamak fusion reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially the first wall and blanket. Type 316 stainless steel in the 20% cold-worked condition has been irradiated in the HFIR in order to introduce helium as well as displacement damage. A miniature hourglass specimen was developed for the reactor irradiations and subsequent fully reversed low cycle fatigue testing. For material irradiated and tested at 430/sup 0/C in vacuum to a damage level of 7 to 15 dpa and containing 200 to 1000 appm He, a reduction in life by amore » factor of 3 to 10 was observed. An attempt was made to predict irradiated fatigue life by fitting data from irradiated material to a power law equation similar to the universal slopes equation and using ductility ratios from tensile tests to modify the equation for irradiated material.« less

Conceptual design of a thermalhydraulic loop for multiple test geometries at supercritical conditions named Supercritical Phenomena Experimental Test Apparatus (SPETA)

NASA Astrophysics Data System (ADS)

Adenariwo, Adepoju

The efficiency of nuclear reactors can be improved by increasing the operating pressure of current nuclear reactors. Current CANDU-type nuclear reactors use heavy water as coolant at an outlet pressure of up to 11.5 MPa. Conceptual SuperCritical Water Reactors (SCWRs) will operate at a higher coolant outlet pressure of 25 MPa. Supercritical water technology has been used in advanced coal plants and its application proves promising to be employed in nuclear reactors. To better understand how supercritical water technology can be applied in nuclear power plants, supercritical water loops are used to study the heat transfer phenomena as it applies to CANDU-type reactors. A conceptual design of a loop known as the Supercritical Phenomena Experimental Apparatus (SPETA) has been done. This loop has been designed to fit in a 9 m by 2 m by 2.8 m enclosure that will be installed at the University of Ontario Institute of Technology Energy Research Laboratory. The loop include components to safely start up and shut down various test sections, produce a heat source to the test section, and to remove reject heat. It is expected that loop will be able to investigate the behaviour of supercritical water in various geometries including bare tubes, annulus tubes, and multi-element-type bundles. The experimental geometries are designed to match the fluid properties of Canadian SCWR fuel channel designs so that they are representative of a practical application of supercritical water technology in nuclear plants. This loop will investigate various test section orientations which are the horizontal, vertical, and inclined to investigate buoyancy effects. Frictional pressure drop effects and satisfactory methods of estimating hydraulic resistances in supercritical fluid shall also be estimated with the loop. Operating limits for SPETA have been established to be able to capture the important heat transfer phenomena at supercritical conditions. Heat balance and flow calculations have been done to appropriately size components in the loop. Sensitivity analysis has been done to find the optimum design for the loop.

Fluid dynamics of the shock wave reactor

NASA Astrophysics Data System (ADS)

Masse, Robert Kenneth

2000-10-01

High commercial incentives have driven conventional olefin production technologies to near their material limits, leaving the possibility of further efficiency improvements only in the development of entirely new techniques. One strategy known as the Shock Wave Reactor, which employs gas dynamic processes to circumvent limitations of conventional reactors, has been demonstrated effective at the University of Washington. Preheated hydrocarbon feedstock and a high enthalpy carrier gas (steam) are supersonically mixed at a temperature below that required for thermal cracking. Temperature recovery is then effected via shock recompression to initiate pyrolysis. The evolution to proof-of-concept and analysis of experiments employing ethane and propane feedstocks are presented. The Shock Wave Reactor's high enthalpy steam and ethane flows severely limit diagnostic capability in the proof-of-concept experiment. Thus, a preliminary blow down supersonic air tunnel of similar geometry has been constructed to investigate recompression stability and (especially) rapid supersonic mixing necessary for successful operation of the Shock Wave Reactor. The mixing capabilities of blade nozzle arrays are therefore studied in the air experiment and compared with analytical models. Mixing is visualized through Schlieren imaging and direct photography of condensation in carbon dioxide injection, and interpretation of visual data is supported by pressure measurement and flow sampling. The influence of convective Mach number is addressed. Additionally, thermal behavior of a blade nozzle array is analyzed for comparison to data obtained in the course of succeeding proof-of-concept experiments. Proof-of-concept is naturally succeeded by interest in industrial adaptation of the Shock Wave Reactor, particularly with regard to issues involving the scaling and refinement of the shock recompression. Hence, an additional, variable geometry air tunnel has been constructed to study the parameter dependence of shock recompression in ducts. Distinct variation of the flow Reynolds and Mach numbers and section height allow unique mapping of each of these parameter dependencies. Agreement with a new one-dimensional model is demonstrated, predicting an exponential pressure profile characterized by two key parameters, the maximum pressure recovery and a characteristic length scale. Transition from one to two-dimensional dependence of the length parameter is observed as the duct aspect ratio varies significantly from unity.

Cost Reductions for Wastewater Treatment Utilizing Water Management at Holston Army Ammunition Plant

DTIC Science & Technology

1976-05-01

says that the granular carbon used is made from bituminous coal. As the waste stream pass through a bed of carbon granules, com- pounds are adsorbed to...findings of laboratory-scale reactor studies conducted at Purdue University for * Clark, Dietz and Associates. The original recommendations and cost...Pretreatment Denitrification by Submerged Anaerbbic I ilters I ~ Trickling Filters S F ,2al Clarification "•’i Pump - ~ Sludge ,Treatment Dual Media Filh:ration

New 5 Kilowatt Free-piston Stirling Space Convertor Developments

NASA Technical Reports Server (NTRS)

Brandhorst, Henry W., Jr.; Chapman, Peter A., Jr.

2007-01-01

The NASA Vision for Exploration of the moon may someday require a nuclear reactor coupled with a free-piston Stirling convertor at a power level of 30-40 kW. In the 1990s, Mechanical Technology Inc. s Stirling Engine Systems Division (some of whose Stirling personnel are now at Foster-Miller, Inc.) developed a 25 kW free piston Stirling Space Power Demonstrator Engine under the SP-100 program. This system consisted of two 12.5 kW engines connected at their hot ends and mounted in tandem to cancel vibration. Recently, NASA and DoE have been developing dual 55 W and 80 W Stirling convertor systems for potential use with radioisotope heat sources. Total test times of all convertors in this effort exceed 120,000 hours. Recently, NASA began a new project with Auburn University to develop a 5 kW, single convertor for potential use in a lunar surface reactor power system. Goals of this development program include a specific power in excess of 140 W/kg at the convertor level, lifetime in excess of five years and a control system that will safely manage the convertors in case of an emergency. Auburn University awarded a subcontract to Foster-Miller, Inc. to undertake development of the 5 kW Stirling Convertor Assembly. The characteristics of the design along with progress in developing the system will be described.

Ion source development for a photoneutralization based NBI system for fusion reactors

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

Simonin, A.; Esch, H. P. L. de; Garibaldi, P.

2015-04-08

The next step after ITER is to demonstrate the viability and generation of electricity by a future fusion reactor (DEMO). The specifications required to operate an NBI system on DEMO are very demanding. The system has to provide a very high level of power and energy, ~100MW of D° beam at 1MeV, including high wall-plug efficiency (η > 60%). For this purpose, a new injector concept, called Siphore, is under investigation between CEA and French universities. Siphore is based on the stripping of the accelerated negative ions by photo-detachment provided by several Fabry-Perot cavities (3.5MW of light power per cavity)more » implemented along the D{sup −} beam. The beamline is designed to be tall and narrow in order that the photon flux overlaps the entire negative ion beam. The paper will describe the present R and D at CEA which addresses the development of an ion source and pre-accelerator prototypes for Siphore, the main goal being to produce an intense negative ion beam sheet. The negative ion source Cybele is based on a magnetized plasma column where hot electrons are emitted from the source center. Parametric studies of the source are performed using Langmuir probes in order to characterize the plasma and to compare with numerical models being developed in French universities.« less

Neutron elastic and inelastic cross section measurements for 28Si

NASA Astrophysics Data System (ADS)

Derdeyn, E. C.; Lyons, E. M.; Morin, T.; Hicks, S. F.; Vanhoy, J. R.; Peters, E. E.; Ramirez, A. P. D.; McEllistrem, M. T.; Mukhopadhyay, S.; Yates, S. W.

2017-09-01

Neutron elastic and inelastic cross sections are critical for design and implementation of nuclear reactors and reactor equipment. Silicon, an element used abundantly in fuel pellets as well as building materials, has little to no experimental cross sections in the fast neutron region to support current theoretical evaluations, and thus would benefit from any contribution. Measurements of neutron elastic and inelastic differential scattering cross sections for 28Si were performed at the University of Kentucky Accelerator Laboratory for incident neutron energies of 6.1 MeV and 7.0 MeV. Neutrons were produced by accelerated deuterons incident on a deuterium gas cell. These nearly mono-energetic neutrons then scattered off a natural Si sample and were detected using liquid deuterated benzene scintillation detectors. Scattered neutron energy was deduced using time-of-flight techniques in tandem with kinematic calculations for an angular distribution. The relative detector efficiency was experimentally determined over a neutron energy range from approximately 0.5 to 7.75 MeV prior to the experiment. Yields were corrected for multiple scattering and neutron attenuation in the sample using the forced-collision Monte Carlo correction code MULCAT. Resulting cross sections will be presented along with comparisons to various data evaluations. Research is supported by USDOE-NNSA-SSAP: NA0002931, NSF: PHY-1606890, and the Donald A. Cowan Physics Institute at the University of Dallas.

Plasmolysis for efficient CO2 -to-fuel conversion

NASA Astrophysics Data System (ADS)

van Rooij, Gerard

2015-09-01

The strong non-equilibrium conditions provided by the plasma phase offer the opportunity to beat traditional thermal process energy efficiencies via preferential excitation of molecular vibrational modes. It is therefore a promising option for creating artificial solar fuels from CO2as raw material using (intermittently available) sustainable energy surpluses, which can easily be deployed within the present infrastructure for conventional fossil fuels. In this presentation, a common microwave reactor approach is evaluated experimentally with Rayleigh scattering and Fourier transform infrared spectroscopy to assess gas temperatures and conversion degrees, respectively. The results are interpreted on basis of estimates of the plasma dynamics obtained with electron energy distribution functions calculated with a Boltzmann solver. It indicates that the intrinsic electron energies are higher than is favourable for preferential vibrational excitation due to dissociative excitation, which causes thermodynamic equilibrium chemistry still to dominate the initial experiments. Novel reactor approaches are proposed to tailor the plasma dynamics to achieve the non-equilibrium in which vibrational excitation is dominant. In collaboration with Dirk van den Bekerom, Niek den Harder, Teofil Minea, Dutch Institute For Fundamental Energy Research, Eindhoven, Netherlands; Gield Berden, Institute for Molecules and Materials, FELIX facility, Radboud University, Nijmegen, Netherlands; Richard Engeln, Applied Physics, Plasma en Materials Processing, Eindhoven University of Technology; and Waldo Bongers, Martijn Graswinckel, Erwin Zoethout, Richard van de Sanden, Dutch Institute For Fundamental Energy Research, Eindhoven, Netherlands.

New 5 kW free-piston Stirling space convertor developments

NASA Astrophysics Data System (ADS)

Brandhorst, Henry W., Jr.; Chapman, Peter A., Jr.

2008-07-01

The NASA Vision for Exploration of the moon may someday require a nuclear reactor coupled with a free-piston Stirling convertor at a power level of 30-40 kW. In the 1990s, Mechanical Technology Inc.'s Stirling Engine Systems Division (some of whose Stirling personnel are now at Foster-Miller, Inc.) developed a 25 kW free-piston Stirling Space Power Demonstrator Engine under the SP-100 program. This system consisted of two 12.5 kW engines connected at their hot ends and mounted in tandem to cancel vibration. Recently, NASA and DoE have been developing dual 55 and 80 W Stirling convertor systems for potential use with radioisotope heat sources. Total test times of all convertors in this effort exceed 120,000 h. Recently, NASA began a new project with Auburn University to develop a 5 kW, single convertor for potential use in a lunar surface reactor power system. Goals of this development program include a specific power in excess of 140 W/kg at the convertor level, lifetime in excess of five years and a control system that will safely manage the convertors in case of an emergency. Auburn University awarded a subcontract to Foster-Miller, Inc. to undertake development of the 5 kW Stirling convertor assembly. The characteristics of the design along with progress in developing the system will be described.

A simulator-based nuclear reactor emergency response training exercise.

PubMed

Waller, Edward; Bereznai, George; Shaw, John; Chaput, Joseph; Lafortune, Jean-Francois

Training offsite emergency response personnel basic awareness of onsite control room operations during nuclear power plant emergency conditions was the primary objective of a week-long workshop conducted on a CANDU® virtual nuclear reactor simulator available at the University of Ontario Institute of Technology, Oshawa, Canada. The workshop was designed to examine both normal and abnormal reactor operating conditions, and to observe the conditions in the control room that may have impact on the subsequent offsite emergency response. The workshop was attended by participants from a number of countries encompassing diverse job functions related to nuclear emergency response. Objectives of the workshop were to provide opportunities for participants to act in the roles of control room personnel under different reactor operating scenarios, providing a unique experience for participants to interact with the simulator in real-time, and providing increased awareness of control room operations during accident conditions. The ability to "pause" the simulator during exercises allowed the instructors to evaluate and critique the performance of participants, and to provide context with respect to potential offsite emergency actions. Feedback from the participants highlighted (i) advantages of observing and participating "hands-on" with operational exercises, (ii) their general unfamiliarity with control room operational procedures and arrangements prior to the workshop, (iii) awareness of the vast quantity of detailed control room procedures for both normal and transient conditions, and (iv) appreciation of the increased workload for the operators in the control room during a transient from normal operations. Based upon participant feedback, it was determined that the objectives of the training had been met, and that future workshops should be conducted.

The influence of solid retention time on IFAS-MBR systems: Assessment of nitrous oxide emission.

PubMed

Mannina, Giorgio; Capodici, Marco; Cosenza, Alida; Laudicina, Vito Armando; Di Trapani, Daniele

2017-12-01

The aim of the present study was to investigate the nitrous oxide (N 2 O) emissions from a moving bed based Integrated Fixed Film Activated Sludge (IFAS) - membrane bioreactor (MBR) pilot plant, designed according to the University of Cape Town (UCT) layout. The experimental campaign had a duration of 110 days and was characterized by three different sludge retention time (SRT) values (∞, 30 d and 15 d). Results highlighted that N 2 O concentrations decreased when the biofilm concentrations increased within the aerobic reactor. Results have shown an increase of N 2 O with the decrease of SRT. Specifically, an increase of N 2 O-N emission factor occurred with the decrease of the SRT (0.13%, 0.21% and 0.76% of influent nitrogen for SRT = ∞, SRT = 30 d and SRT = 15 d, respectively). Moreover, the MBR tank resulted the key emission source (up to 70% of the total N 2 O emission during SRT = ∞ period) whereas the highest N 2 O production occurred in the anoxic reactor. Moreover, N 2 O concentrations measured in the permeate flow were not negligible, thus highlighting its potential detrimental contribution for the receiving water body. The role of each plant reactor as N 2 O-N producer/consumer varies with the SRT variation, indeed the aerobic reactor was a N 2 O consumer at SRT = ∞ and a producer at SRT = 30 d. Copyright © 2017 Elsevier Ltd. All rights reserved.

The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

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

Garrison, L. M., E-mail: garrisonlm@ornl.gov; Egle, B. J.; Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706

2016-08-15

The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ionmore » gun can irradiate the samples with ion currents of 20 μA–500 μA; the typical current used is 72 μA, which is an average flux of 9 × 10{sup 14} ions/(cm{sup 2} s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.« less

[Bacterial diversity in sequencing batch biofilm reactor (SBBR) for landfill leachate treatment using PCR-DGGE].

PubMed

Xiao, Yong; Yang, Zhao-hui; Zeng, Guang-ming; Ma, Yan-he; Liu, You-sheng; Wang, Rong-juan; Xu, Zheng-yong

2007-05-01

For studying the bacterial diversity and the mechanism of denitrification in sequencing bath biofilm reactor (SBBR) treating landfill leachate to provide microbial evidence for technique improvements, total microbial DNA was extracted from samples which were collected from natural landfill leachate and biofilm of a SBBR that could efficiently remove NH4+ -N and COD of high concentration. 16S rDNA fragments were amplified from the total DNA successfully using a pair of universal bacterial 16S rDNA primer, GC341F and 907R, and then were used for denaturing gradient gel electrophoresis (DGGE) analysis. The bands in the gel were analyzed by statistical methods and excided from the gel for sequencing, and the sequences were used for homology analysis and then two phylogenetic trees were constructed using DNAStar software. Results indicated that the bacterial diversity of the biofilm in SBBR and the landfill leachate was abundant, and no obvious change of community structure happened during running in the biofilm, in which most bacteria came from the landfill leachate. There may be three different modes of denitrification in the reactor because several different nitrifying bacteria, denitrifying bacteria and anaerobic ammonia oxidation bacteria coexisted in it. The results provided some valuable references for studying microbiological mechanism of denitrification in SBBR.

The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

NASA Astrophysics Data System (ADS)

Scarlat, Raluca O.; Peterson, Per F.

2014-01-01

The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

Development of attrition resistant iron-based Fischer-Tropsch catalysts

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

NONE

2000-09-20

The Fischer-Tropsch (F-T) reaction provides a way of converting coal-derived synthesis gas (CO+H{sub 2}) to liquid fuels. Since the reaction is highly exothermic, one of the major problems in control of the reaction is heat removal. Recent work has shown that the use of slurry bubble column reactors (SBCRs) can largely solve this problem. The use of iron-based catalysts is attractive not only due to their low cost and ready availability, but also due to their high water-gas shift activity which makes it possible to use these catalysts with low H{sub 2}/CO ratios. However, a serious problem with use ofmore » Fe catalysts in a SBCR is their tendency to undergo attrition. This can cause fouling/plugging of downstream filters and equipment, makes the separation of catalyst from the oil/wax product very difficult if not impossible, and results a steady loss of catalyst from the reactor. The objective of this research is to develop robust iron-based Fischer-Tropsch catalysts that have suitable activity, selectivity and stability to be used in the slurry bubble column reactor. Specifically we aim to develop to: (1) improve the performance and preparation procedure of the high activity, high attrition resistant, high alpha iron-based catalysts synthesized at Hampton University (2) seek improvements in the catalyst performance through variations in process conditions, pretreatment procedures and/or modifications in catalyst preparation steps and (3) investigate the performance in a slurry reactor. The effort during the reporting period has been devoted to effects of pretreating procedures, using H{sub 2}, CO and syngas (H{sub 2}/CO = 0.67) as reductants, on the performance (activity, selectivity and stability with time) of a precipitated iron catalyst (100Fe/5Cu/4.2K/10SiO{sub 2} on a mass basis ) during F-T synthesis were studied in a fixed-bed reactor.« 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

Thermal neutron streaming effects and WIMS analysis of the Penn State subcritical graphite pile

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

Feltus, M.A.; Zediak, C.S.; Jester, W.A.

1997-12-01

This analysis was performed on the Pennsylvania State University (PSU) subcritical reactor to find more accurate values for such nuclear parameters as the thermal fuel utilization factor, thermal diffusion length in the graphite, migration area, k{sub eff}, etc. The analysis involved using the Winfrith Integrated Multigroup Scheme (WIMS) code as well as various hand calculations to find and compare those parameters. The data found in this analysis will be used by future students in the Penn State laboratory courses.

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

Saleh, Tarik A.; Quintana, Matthew Estevan; Romero, Tobias J.

As a part of the project “High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation” an Integrated Research Program (IRP) project from the U.S. Department of Energy, Nuclear Energy University Programs (NEUP), TEM geometry samples of ferritic cladding alloys, Ni based super alloys and model alloys were irradiated in the BOR-60 reactor to ~16 dpa at ~370°C and ~400°C. Samples were sent to Los Alamos National Laboratory and subjected to shear punch testing. This report presents the results from this testing.

Treatment of concentrated industrial wastewaters originating from oil shale and the like by electrolysis polyurethane foam interaction

DOEpatents

Tiernan, Joan E.

1991-01-01

Highly concentrated and toxic petroleum-based and synthetic fuels wastewaters such as oil shale retort water are treated in a unit treatment process by electrolysis in a reactor containing oleophilic, ionized, open-celled polyurethane foams and subjected to mixing and l BACKGROUND OF THE INVENTION The invention described herein arose in the course of, or under, Contract No. DE-AC03-76SF00098 between the U.S. Department of Energy and the University of California.

The Serpent Monte Carlo Code: Status, Development and Applications in 2013

NASA Astrophysics Data System (ADS)

Leppänen, Jaakko; Pusa, Maria; Viitanen, Tuomas; Valtavirta, Ville; Kaltiaisenaho, Toni

2014-06-01

The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.

Decision Support For Digester Algae Integration For Improved Environmental And Economic Sustainability

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

Guillen, Donna P.; Panike, Katherine R.; Havlovick, Caryn M.

The Idaho National Laboratory (INL) has teamed with University of Idaho and Boise State University to make the use of ADs more attractive by implementing a two-stage AD and coupling additional processes to the system. The addition of a polyhydroxyalkanoate (PHA) reactor, algae cultivation system, and a biomass treatment system such as fast-pyrolysis or hydrothermal liquefaction (HTL) would further sequester carbon and nutrients, as well as add valuable products that can be sold or used on-site to mitigate costs. The Decision-support for Digester-Algae IntegRation for Improved Environmental and Economic Sustainability (DAIRIEES) technoeconomic model will play a key role in evaluatingmore » the effectiveness and viability of this system to achieve economic and environmental sustainability by the dairy industry.« less

Spectral Performance of a Composite Single-Crystal Filtered Thermal Neutron Beam for BNCT Research at the University of Missouri

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

J. Brockman; D. W. Nigg; M. F. Hawthorne

2009-07-01

Parameter studies, design calculations and initial neutronic performance measurements have been completed for a new thermal neutron beamline to be used for neutron capture therapy cell and small-animal radiobiology studies at the University of Missouri Research Reactor. The beamline features the use of single-crystal silicon and bismuth sections for neutron filtering and for reduction of incident gamma radiation. The calculated and measured thermal neutron fluxes produced at the irradiation location are 9.6x108 and 8.8x108 neutrons/cm2-s, respectively. Calculated and measured cadmium ratios (Au foils) are 217 and 132. These results indicate a well-thermalized neutron spectrum with sufficient thermal neutron flux formore » a variety of small animal BNCT studies.« less

Innovations in Nuclear Infrastructure and Education

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

John Bernard

The decision to implement the Innovation in Nuclear Infrastructure and Engineering Program (INIE) was an important first step towards ensuring that the United States preserves its worldwide leadership role in the field of nuclear science and engineering. Prior to INIE, university nuclear science and engineering programs were waning, undergraduate student enrollment was down, university research reactors were being shut down, while others faced the real possibility of closure. For too long, cutting edge research in the areas of nuclear medicine, neutron scattering, radiochemistry, and advanced materials was undervalued and therefore underfunded. The INIE program corrected this lapse in focus andmore » direction and started the process of drawing a new blueprint with positive goals and objectives that supports existing as well the next generation of educators, students and researchers.« less

Human capital needs - teaching, training and coordination for nuclear fuel cycle

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

Retegan, T.; Ekberg, C.; John, J.

Human capital is the accumulation of competencies, knowledge, social and creativity skills and personality attributes, which are necessary to perform work so as to produce economic value. In the frame of the nuclear fuel cycle, this is of paramount importance that the right human capital exists and in Europe this is fostered by a series of integrated or directed projects. The teaching, training and coordination will be discussed in the frame of University curricula with examples from several programs, like e.g. the Master of Nuclear Engineering at Chalmers University, Sweden and two FP7 EURATOM Projects: CINCH - a project formore » cooperation in nuclear chemistry - and ASGARD - a research project on advanced or novel nuclear fuels and their reprocessing issues for generation IV reactors. The integration of the university curricula in the market needs but also the anchoring in the research and future fuel cycles will be also discussed, with examples from the ASGARD project. (authors)« less

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

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

Woods, Brian; Gutowska, Izabela; Chiger, Howard

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

Boron neutron capture therapy induces apoptosis of glioma cells through Bcl-2/Bax

PubMed Central

2010-01-01

Background Boron neutron capture therapy (BNCT) is an alternative treatment modality for patients with glioma. The aim of this study was to determine whether induction of apoptosis contributes to the main therapeutic efficacy of BNCT and to compare the relative biological effect (RBE) of BNCT, γ-ray and reactor neutron irradiation. Methods The neutron beam was obtained from the Xi'an Pulsed Reactor (XAPR) and γ-rays were obtained from [60Co] γ source of the Fourth Military Medical University (FMMU) in China. Human glioma cells (the U87, U251, and SHG44 cell lines) were irradiated by neutron beams at the XAPR or [60Co] γ-rays at the FMMU with different protocols: Group A included control nonirradiated cells; Group B included cells treated with 4 Gy of [60Co] γ-rays; Group C included cells treated with 8 Gy of [60Co] γ-rays; Group D included cells treated with 4 Gy BPA (p-borono-phenylalanine)-BNCT; Group E included cells treated with 8 Gy BPA-BNCT; Group F included cells irradiated in the reactor for the same treatment period as used for Group D; Group G included cells irradiated in the reactor for the same treatment period as used for Group E; Group H included cells irradiated with 4 Gy in the reactor; and Group I included cells irradiated with 8 Gy in the reactor. Cell survival was determined using the 3-(4,5-dimethylthiazol-2-yl-2,5-diphenyltetrazolium (MTT) cytotoxicity assay. The morphology of cells was detected by Hoechst33342 staining and transmission electron microscope (TEM). The apoptosis rate was detected by flow cytometer (FCM). The level of Bcl-2 and Bax protein was measured by western blot analysis. Results Proliferation of U87, U251, and SHG44 cells was much more strongly inhibited by BPA-BNCT than by irradiation with [60Co] γ-rays (P < 0.01). Nuclear condensation was determined using both a fluorescence technique and electron microscopy in all cell lines treated with BPA-BNCT. Furthermore, the cellular apoptotic rates in Group D and Group E treated with BPA-BNCT were significantly higher than those in Group B and Group C irradiated by [60Co] γ-rays (P < 0.01). The clonogenicity of glioma cells was reduced by BPA-BNCT compared with cells treated in the reactor (Group F, G, H, I), and with the control cells (P < 0.01). Upon BPA-BNCT treatment, the Bax level increased in glioma cells, whereas Bcl-2 expression decreased. Conclusions Compared with γ-ray and reactor neutron irradiation, a higher RBE can be achieved upon treatment of glioma cells with BNCT. Glioma cell apoptosis induced by BNCT may be related to activation of Bax and downregulation of Bcl-2. PMID:21122152

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

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

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

2010-09-01

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

Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

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

Zhang, Hongbin; Szilard, Ronaldo; Epiney, Aaron

Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance duringmore » LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.« less

Transport of Neutrons and Photons Through Iron and Water Layers

NASA Astrophysics Data System (ADS)

Košťál, Michal; Cvachovec, František; Ošmera, Bohumil; Hansen, Wolfgang; Noack, Klaus

2009-08-01

The neutron and photon spectra were measured after iron and water plates placed at the horizontal channel of the Dresden University reactor AK-2. The measurements have been performed with the multiparameter spectrometer [1] with a stilbene cylindrical crystal, 10 × 10 mm or 45 × 45 mm; the neutron and photon spectra have been measured simultaneously. The calculations were performed with the MCNP code and nuclear data libraries ENDF/B VI.2, ENDF/BVII.0, JENDL 3.3 and JEFF 3.1. The measured channel leakage spectrum was used as the input spectrum for the transport calculation. Photons, the primary photons from the reactor - as well as the ones induced by neutron interaction - were calculated. The comparison of the measurements and calculations through 10 cm of iron and 20 cm thickness of water are presented. Besides that, the attenuation of the radiation mixed field by iron layers from 5 to 30 cm is presented; the measured and calculated data are compared.

Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

NASA Astrophysics Data System (ADS)

Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

2018-02-01

The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

Diode laser sensor to monitor HCL in a plasma etch reactor

NASA Astrophysics Data System (ADS)

Kim, Suhong; Klimecky, Pete; Chou, Shang-I.; Jeffries, Jay B.; Terry, Fred L., Jr.; Hanson, Ronald K.

2002-09-01

Absorption measurements of HCl during plasma etching of poly-silicon are made using the P(4) transition in the first vibrational overtone band near 1.79 μm. Single path absorption provides a real-time HCl monitor during etching of six-inch wafers in a commercial Lam Research 9400SE reactor at the University of Michigan. Wavelength modulation at 10.7 MHz is used to distinguish the absorption signal from the strong plasma emission. The laser center frequency is ramp-tuned at 500 Hz providing an HCl measurement every 2ms. Direct absorption measurements without the plasma are used to calibrate the wavelength modulation signal. The minimum detectable absorbance was 5x(10)-6 with 50 ms averaging, leading to an HCl detection limit of ~(10)12cm-3. For a given ratio of the feedstock HBr/Cl2, the measured HCl concentration tracks the average etch rate. These measurements demonstrate the feasibility of a real-time diode laser-based etch rate sensor.

GoAmazon2014/15. Oxidation Flow Reactor Final Campaign Report

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

Jimenez, J. L.; Day, D. A.; Hu, W.

The primary goal of the Green Ocean Amazon (GoAmazon2014/5) field campaign was to measure and mechanistically understand the formation of particle number and mass in a region affected by large tropical rainforest biogenic emissions and sometimes anthropogenic influence from a large urban center. As part of the two intensive operational periods (IOPs) and in collaboration with Pacific Northwest National Laboratory (PNNL) and Harvard, the Jimenez Group proposed to deploy a High-Resolution Time-of-Flight Aerosol Mass Spectrometer (HR-ToF-AMS), Thermal Denuder (TD), Scanning Mobility Particle Size (SMPS), two oxidation flow reactors (OFR; including supporting O 3, CO/CO 2/CH 4, RH analyzers), and amore » high volume filter sampler (MCV) for the measurement of gas and aerosol chemical, physicochemical, and volatility properties. The two IOPs were conducted during the wet season (February to March, 2014) and dry season (August to October, 2014). This proposal was part of a collaborative proposal involving other university and government laboratories.« less

Irradiated Beryllium Disposal Workshop, Idaho Falls, ID, May 29-30, 2002

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

Longhurst, Glen Reed; Anderson, Gail; Mullen, Carlan K

2002-07-01

In 2001, while performing routine radioactive decay heat rate calculations for beryllium reflector blocks for the Advanced Test Reactor (ATR), it became evident that there may be sufficient concentrations of transuranic isotopes to require classification of this irradiated beryllium as transuranic waste. Measurements on samples from ATR reflector blocks and further calculations confirmed that for reflector blocks and outer shim control cylinders now in the ATR canal, transuranic activities are about five times the threshold for classification. That situation implies that there is no apparent disposal pathway for this material. The problem is not unique to the ATR. The Highmore » Flux Isotope Reactor at Oak Ridge National Laboratory, the Missouri University Research Reactor at Columbia, Missouri and other reactors abroad must also deal with this issue. A workshop was held in Idaho Falls Idaho on May 29-30, 2002 to acquaint stakeholders with these findings and consider a path forward in resolving the issues attendant to disposition of irradiated material. Among the findings from this workshop were (1) there is a real potential for the US to be dependent on foreign sources for metallic beryllium within about a decade; (2) there is a need for a national policy on beryllium utilization and disposition and for a beryllium coordinating committee to be assembled to provide guidance on that policy; (3) it appears it will be difficult to dispose of this material at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico due to issues of Defense classification, facility radioactivity inventory limits, and transportation to WIPP; (4) there is a need for a funded DOE program to seek resolution of these issues including research on processing techniques that may make this waste acceptable in an existing disposal pathway or allow for its recycle.« less

Development and Assessment of CTF for Pin-resolved BWR Modeling

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

Salko, Robert K; Wysocki, Aaron J; Collins, Benjamin S

2017-01-01

CTF is the modernized and improved version of the subchannel code, COBRA-TF. It has been adopted by the Consortium for Advanced Simulation for Light Water Reactors (CASL) for subchannel analysis applications and thermal hydraulic feedback calculations in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). CTF is now jointly developed by Oak Ridge National Laboratory and North Carolina State University. Until now, CTF has been used for pressurized water reactor modeling and simulation in CASL, but in the future it will be extended to boiling water reactor designs. This required development activities to integrate the code into the VERA-CSmore » workflow and to make it more ecient for full-core, pin resolved simulations. Additionally, there is a significant emphasis on producing high quality tools that follow a regimented software quality assurance plan in CASL. Part of this plan involves performing validation and verification assessments on the code that are easily repeatable and tied to specific code versions. This work has resulted in the CTF validation and verification matrix being expanded to include several two-phase flow experiments, including the General Electric 3 3 facility and the BWR Full-Size Fine Mesh Bundle Tests (BFBT). Comparisons with both experimental databases is reasonable, but the BFBT analysis reveals a tendency of CTF to overpredict void, especially in the slug flow regime. The execution of these tests is fully automated, analysis is documented in the CTF Validation and Verification manual, and the tests have become part of CASL continuous regression testing system. This paper will summarize these recent developments and some of the two-phase assessments that have been performed on CTF.« less

Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

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

Petrovic, Bojan; Maldonado, Ivan

2016-04-14

The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed onmore » December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.« less

PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

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

Smith, P.; Sheetz, S.

Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply. The primary module in the model compares the consumption of uranium for various production reactors against the base case ofmore » Watts Bar I running a nominal load of 1,696 tritium producing burnable absorber rods (TPBARs) with an average refueling of 41,000 kg low enriched uranium (LEU) on an 18 month cycle. After inputting an initial year, starting inventory of unobligated uranium and tritium production forecast, the model will compare and contrast the depletion rate of the LEU between the entered alternatives. This is an annual tritium production rate of approximately 0.059 grams of tritium per kilogram of LEU (g-T/kg-LEU). To date, the Nuclear Regulatory Commission (NRC) license has not been amended to accept a full load of TPBARs so the nominal tritium production has not yet been achieved. The alternatives currently loaded into the model include the three light water SMRs evaluated in TAMU senior projects including, mPower, Holtec and NuScale designs. Initial evaluations of tritium production in light water reactor (LWR) based SMRs using optimized loads TPBARs is on the order 0.02-0.06 grams of tritium per kilogram of LEU used. The TAMU students also chose to model tritium production in the GE-Hitachi SPRISM, a pooltype sodium fast reactor (SFR) utilizing a modified TPBAR type target. The team was unable to complete their project so no data is available. In order to include results from a fast reactor, the SRNL Technical Advisory Committee (TAC) ran a Monte Carlo N-Particle (MCNP) model of a basic SFR for comparison. A 600MWth core surrounded by a lithium blanket produced approximately 1,000 grams of tritium annually with a 13% enriched, 6 year core. This is similar results to a mid-1990’s study where the Fast Flux Test Facility (FFTF), a 400 MWth reactor at the Idaho National Laboratory (INL), could produce about 1,000 grams with an external lithium target. Normalized to the LWRs values, comparative tritium production for an SFR could be approximately 0.31 g-T/kg LEU.« less

The Development of Neutron Radiography and Tomography on a SLOWPOKE-2 Reactor

NASA Astrophysics Data System (ADS)

Bennett, L. G. I.; Lewis, W. J.; Hungler, P. C.

Development of neutron radiography at the Royal Military College of Canada (RMC) started by trying to interest the Royal Canadian Air Force (RCAF) in this new non-destructive testing (NDT) technique. A Californium-252 based device was ordered and then installed at RMC for development of applicable techniques for aircraft by the first author. A second and transportable device was then designed, modified and used in trials at RCAF Bases and other locations for one year. This activity was the only foreign loan of the U.S. Californium Loan Program. Around this time, SLOWPOKE-2 reactors were being installed at four Canadian universities, while a new science and engineering building was being built at RMC. A reactor pool was incorporated and efforts to procure a reactor succeeded a decade later with a SLOWPOKE-2 reactor being installed at RMC. The only modification by the vendor for RMC was a thermal column replacing an irradiation site inside the reactor container for a later installation of a neutron beam tube (NBT). Development of a working NBT took several years, starting with the second author. A demonstration of the actual worth of neutron radiography took place with a CF-18 Hornet aircraft being neutron and X-radiographed at McClellan Air Force Base, Sacramento, CA. This inspection was followed by one of the rudders that had indications of water ingress being radiographed successfully at RMC just after the NBT became functional. The next step was to develop a neutron radioscopy system (NRS), initially employing film and then digital imaging, and is in use today for all flight control surfaces (FCS). With the third author, a technique capable of removing water from affected FCS was developed at RMC. Heating equipment and a vacuum system were utilized to carefully remove the water. This technique was proven using a sequence of near real time neutron images obtained during the drying process. The results of the drying process were correlated with a relative humidity gauge and an NDT technique that could be performed at Canadian Forces (CF) Bases was developed. In order to determine the structural integrity of the component having undergone this water removal, further research was required into the effect of water inside composite honeycomb structures. This need has led to the present development of neutron tomography on the reactor at RMC, which is capable of determining the exact location of water ingress inside composite components. This technique has been successfully applied to coupons as well as to complete rudders.

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

NONE

The objective of Task 1 is to prepare and evaluate catalysts and to develop efficient reactor systems for the selective conversion of hydrogen-lean synthesis gas to alcohol fuel extender and octane enhancers. Task 1 is subdivided into three separate subtasks: laboratory and equipment setup; catalysis research; and reaction engineering and modeling. Research at West Virginia University (WVU) is focused on molybdenum-based catalysts for higher alcohol synthesis. Parallel research carried out at Union Carbide Corporation (UCC) is focused on transition-metal-oxide catalysts. During this time period, at WVU, we tried several methods to eliminate problems related to condensation of heavier products whenmore » reduced Mo-Ni-K/C materials were used as catalysts. We then resumed our kinetic study on the reduced Mo-Ni-K/C materials were used as catalysts. We then resumed our kinetic study on the reduced Mo-Ni-K/C catalysts. We have also obtained same preliminary results in our attempts to analyze quantitatively the temperature-programmed reduction spectra for C- supported Mo-based catalysts. We have completed the kinetic study for the sulfided Co-K-MoS{sub 2}/C catalyst. We have compared the results of methanol synthesis using the membrane reactor with those using a simple plug-flow reactor. At UCC, the complete characterization of selected catalysts has been completed. The results suggest that catalyst pretreatment under different reducing conditions yield different surface compositions and thus different catalytic reactivities.« less

Optimization of on-line hydrogen stable isotope ratio measurements of halogen- and sulfur-bearing organic compounds using elemental analyzer-chromium/high-temperature conversion isotope ratio mass spectrometry (EA-Cr/HTC-IRMS).

PubMed

Gehre, Matthias; Renpenning, Julian; Geilmann, Heike; Qi, Haiping; Coplen, Tyler B; Kümmel, Steffen; Ivdra, Natalija; Brand, Willi A; Schimmelmann, Arndt

2017-03-30

Accurate hydrogen isotopic analysis of halogen- and sulfur-bearing organics has not been possible with traditional high-temperature conversion (HTC) because the formation of hydrogen-bearing reaction products other than molecular hydrogen (H 2 ) is responsible for non-quantitative H 2 yields and possible hydrogen isotopic fractionation. Our previously introduced, new chromium-based EA-Cr/HTC-IRMS (Elemental Analyzer-Chromium/High-Temperature Conversion Isotope Ratio Mass Spectrometry) technique focused primarily on nitrogen-bearing compounds. Several technical and analytical issues concerning halogen- and sulfur-bearing samples, however, remained unresolved and required further refinement of the reactor systems. The EA-Cr/HTC reactor was substantially modified for the conversion of halogen- and sulfur-bearing samples. The performance of the novel conversion setup for solid and liquid samples was monitored and optimized using a simultaneously operating dual-detection system of IRMS and ion trap MS. The method with several variants in the reactor, including the addition of manganese metal chips, was evaluated in three laboratories using EA-Cr/HTC-IRMS (on-line method) and compared with traditional uranium-reduction-based conversion combined with manual dual-inlet IRMS analysis (off-line method) in one laboratory. The modified EA-Cr/HTC reactor setup showed an overall H 2 -recovery of more than 96% for all halogen- and sulfur-bearing organic compounds. All results were successfully normalized via two-point calibration with VSMOW-SLAP reference waters. Precise and accurate hydrogen isotopic analysis was achieved for a variety of organics containing F-, Cl-, Br-, I-, and S-bearing heteroelements. The robust nature of the on-line EA-Cr/HTC technique was demonstrated by a series of 196 consecutive measurements with a single reactor filling. The optimized EA-Cr/HTC reactor design can be implemented in existing analytical equipment using commercially available material and is universally applicable for both heteroelement-bearing and heteroelement-free organic-compound classes. The sensitivity and simplicity of the on-line EA-Cr/HTC-IRMS technique provide a much needed tool for routine hydrogen-isotope source tracing of organic contaminants in the environment. Copyright © 2016 John Wiley & Sons, Ltd. Copyright © 2016 John Wiley & Sons, Ltd.

Microbial community variation and functions to excess sludge reduction in a novel gravel contact oxidation reactor.

PubMed

Lin, Shanshan; Jin, Y; Fu, L; Quan, C; Yang, Y S

2009-06-15

Excess biomass produced within the degradation processes of organic pollutants is creating environmental challenges. The gravel contact oxidation reactor (GCOR) filled with crushed stone globular aggregates as carriers, has been demonstrated capable of reducing the excess sludge effectively in some pilot and small-scale engineering studies. In order to evaluate the variation and structure of the microbial community and their functions to excess sludge reduction in GCOR, a conventional activated sludge reactor (ASR) was studied as a comparison. The 16S rDNA library of the universal bacteria was constructed, Shannon's diversity index (H) and Species evenness (E) were calculated with distance-based operational taxonomic unit and richness (DOTUR) for microbial diversity. Real-time quantity PCR and optical microscope were used for absolute bacterial DNA concentration and eukarya identification, respectively. Meanwhile, the suspended solid index in GCOR and ASR was detected for assessing the excess sludge production. The results indicated that the most abundant bacteria in GCOR were those related to the beta-Proteobacteria group, then gamma-Proteobacteria and to Cytophaga-Flexibacter-Bacteriode (CFB). In the ASR samples major bacteria were in the closest match with gamma-Proteobacteria, then beta-Proteobacteria and CFB. Shannon's index (H) was higher (3.41) for diversity of bacteria extracted from the carrier samples in GCOR than that (2.71) from the sludge sample in ASR. Species evenness (E) for the isolates from GCOR and ASR samples was 0.97 and 0.96, respectively. Comparison of the universal bacteria population in GCOR and ASR shows that the total bacterial DNA concentration on the GCOR carriers were 8.98 x 10(5) microg/ microl, twice that in ASR of 4.67 x 10(5) microg/ microl under normal operation of two reactors. But the MLSS in GCOR was only 4.5mg/L, 25 times less than that in ASR of 115.4 mg/L. The most representative eukarya were protozoa both in GCOR (15 no. per 20 ml) and in ASR (15 no. per 20 ml); the next abundant group was attachment plants 10 no. per 20 ml in GCOR and 4 no. per 20 ml in ASR, respectively. Rotifers and copepoda belonging to metazoan were only present in GCOR (8 no. per 20 ml for both rotifers and copepoda). The microbial diversity and population difference both in the GCOR carriers and ASR sludge indicated that the diverse microbes, a large amount of biomass forming longer microbial food chains attached on the carriers may be the main functions for the excess sludge reduction in GCOR.

Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

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

Richard R. Schultz; Paul D. Bayless; Richard W. Johnson

2010-09-01

The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) beganmore » their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is inadequate to permit steady-state operation at reasonable conditions. 4. To enable the HTTF to operate at a more representative steady-state conditions, DOE recently allocated funding via a DOE subcontract to HTTF to permit an OSU infrastructure upgrade such that 2.2 MW will become available for HTTF experiments. 5. Analyses have been performed to study the relationship between HTTF and MHTGR via the hierarchical two-tiered scaling methodology which has been used successfully in the past, e.g., APEX facility scaling to the Westinghouse AP600 plant. These analyses have focused on the relationship between key variables that will be measured in the HTTF to the counterpart variables in the MHTGR with a focus on natural circulation, using nitrogen as a working fluid, and core heat transfer. 6. Both RELAP5-3D and computational fluid dynamics (CD-Adapco’s STAR-CCM+) numerical models of the MHTGR and the HTTF have been constructed and analyses are underway to study the relationship between the reference reactor and the HTTF. The HTTF is presently being designed. It has ¼-scaling relationship to the MHTGR in both the height and the diameter. Decisions have been made to design the reactor cavity cooling system (RCCS) simulation as a boundary condition for the HTTF to ensure that (a) the boundary condition is well defined and (b) the boundary condition can be modified easily to achieve the desired heat transfer sink for HTTF experimental operations.« less

Dissolution of used nuclear fuel using recycled nitric acid

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

Almond, Philip M.; Daniel, Jr., William E.; Rudisill, Tracy S.

An evaluation was performed on the feasibility of using HB-Line anion exchange column waste streams from Alternate Feedstock 2 (AFS-2) processing for the dissolver solution for used nuclear fuel (UNF) processing. The targeted UNF for dissolution using recycled solution are fuels similar to the University of Missouri Research Reactor (MURR) fuel. Furthermore, the objectives of this experimental program were to validate the feasibility of using impure dissolver solutions with the MURR dissolution flowsheet to verify they would not significantly affect dissolution of the UNF in a detrimental manner.

Computational investigation of suitable polymer gel composition for the QA of the beam components of a BNCT irradiation field.

PubMed

Tanaka, Kenichi; Sakurai, Yoshinori; Hayashi, Shin-Ichiro; Kajimoto, Tsuyoshi; Uchida, Ryohei; Tanaka, Hiroki; Takata, Takushi; Bengua, Gerard; Endo, Satoru

2017-09-01

This study investigated the optimum composition of the MAGAT polymer gel which is to be used in the quality assurance measurement of the thermal neutron, fast neutron and gamma ray components in the irradiation field used for boron neutron capture therapy at the Kyoto University Reactor. Simulations using the PHITS code showed that when combined with the gel, 6 Li concentrations of 0, 10 and 100ppm were found to be potentially usable. Copyright © 2017 Elsevier Ltd. All rights reserved.

Experimental Studies of Coal and Biomass Fuel Synthesis and Flame Characterization for Aircraft Engines

DTIC Science & Technology

2010-05-31

near the catalyst bed was about 600 °C. The rest of the reactor, tar/char collector , and transfer hose to the entrance of the condenser were heated to...such as solar and nuclear [2.2-2.3], to reduce carbon dioxide concentrations and to provide heat for the reactions. Carbon dioxide emission may be...assembly Page 27 of 50 AFOSR Grant #FA9 5 5 0-08-1-04 5 6 Final Performance Report Purdue University Ash collector ( pipe end cap) with product aas

Optimization of on-line hydrogen stable isotope ratio measurements of halogen- and sulfur-bearing organic compounds using elemental analyzer–chromium/high-temperature conversion isotope ratio mass spectrometry (EA-Cr/HTC-IRMS)

USGS Publications Warehouse

Gehre, Matthias; Renpenning, Julian; Geilmann, Heike; Qi, Haiping; Coplen, Tyler B.; Kümmel, Steffen; Ivdra, Natalija; Brand, Willi A.; Schimmelmann, Arndt

2017-01-01

Conclusions: The optimized EA-Cr/HTC reactor design can be implemented in existing analytical equipment using commercially available material and is universally applicable for both heteroelement-bearing and heteroelement-free organic-compound classes. The sensitivity and simplicity of the on-line EA-Cr/HTC-IRMS technique provide a much needed tool for routine hydrogen-isotope source tracing of organic contaminants in the environment. Copyright © 2016 John Wiley & Sons, Ltd.

Dissolution of used nuclear fuel using recycled nitric acid

DOE PAGES

Almond, Philip M.; Daniel, Jr., William E.; Rudisill, Tracy S.

2017-03-20

An evaluation was performed on the feasibility of using HB-Line anion exchange column waste streams from Alternate Feedstock 2 (AFS-2) processing for the dissolver solution for used nuclear fuel (UNF) processing. The targeted UNF for dissolution using recycled solution are fuels similar to the University of Missouri Research Reactor (MURR) fuel. Furthermore, the objectives of this experimental program were to validate the feasibility of using impure dissolver solutions with the MURR dissolution flowsheet to verify they would not significantly affect dissolution of the UNF in a detrimental manner.

NEUTRON RADIATION DAMAGE IN CCD CAMERAS AT JOINT EUROPEAN TORUS (JET).

PubMed

Milocco, Alberto; Conroy, Sean; Popovichev, Sergey; Sergienko, Gennady; Huber, Alexander

2017-10-26

The neutron and gamma radiations in large fusion reactors are responsible for damage to charged couple device (CCD) cameras deployed for applied diagnostics. Based on the ASTM guide E722-09, the 'equivalent 1 MeV neutron fluence in silicon' was calculated for a set of CCD cameras at the Joint European Torus. Such evaluations would be useful to good practice in the operation of the video systems. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

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

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

Finley, V.L. and Levine, J.D.

The results of the 1997 environmental surveillance and monitoring program for the Princeton Plasma Physics Laboratory (PPPL) are presented and discussed. The purpose of this report is to provide the U.S. Department of Energy and the public with information on the level of radioactive and non-radioactive pollutants, if any, that are added to the environment as a result of PPPL's operations. During Calendar Year 1997, PPPL's Tokamak Fusion Test Reactor (TFTR) completed fifteen years of fusion experiments begun in 1982. Over the course of three and half years of deuterium-tritium (D-T) plasma experiments, PPPL set a world record of 10.7more » million watts of controlled fusion power, more than 700 tritium shots pulsed into the reactor vessel generating more than 5.6 x 10 20 neutron and 1.6 gigajoules of fusion energy and researchers studied plasma science experimental data, which included "enhanced reverse shear techniques." As TFTR was completing its historic operations, PPPL participated with the Oak Ridge National Laboratory, Columbia University, and the University of Washington (Seattle) in a collaboration effort to design the National Spherical Torus Experiment (NSTX). This next device, NSTX, is located in the former TFTR Hot Cell on D site, and it is designed to be a smaller and more economical torus fusion reactor. Construction of this device began in late 1997, and first plasma in scheduled for early 1999. For 1997, the U.S. Department of Energy in its Laboratory Appraisal report rated the overall performance of Princeton Plasma Physics Laboratory as "excellent." The report cited the Laboratory's consistently excellent scientific and technological achievements and its successful management practices, which included high marks for environmental management, employee health and safety, human resources administration, science education, and communications. Groundwater investigations continued under a voluntary agreement with the New Jersey Department of Environmental Protection. PPPL monitored the presence of non-radiological contaminants, mainly volatile organic compounds (components of degreasing solvents). Monitoring revealed the presence of low levels of volatile organic compounds in an adjacent area to PPPL. Also, PPPL's radiological monitoring program characterized the ambient, background levels of tritium in the environment and from the TFTR stack; the data are presented in this report.« less

Hardware accelerated high performance neutron transport computation based on AGENT methodology

NASA Astrophysics Data System (ADS)

Xiao, Shanjie

The spatial heterogeneity of the next generation Gen-IV nuclear reactor core designs brings challenges to the neutron transport analysis. The Arbitrary Geometry Neutron Transport (AGENT) AGENT code is a three-dimensional neutron transport analysis code being developed at the Laboratory for Neutronics and Geometry Computation (NEGE) at Purdue University. It can accurately describe the spatial heterogeneity in a hierarchical structure through the R-function solid modeler. The previous version of AGENT coupled the 2D transport MOC solver and the 1D diffusion NEM solver to solve the three dimensional Boltzmann transport equation. In this research, the 2D/1D coupling methodology was expanded to couple two transport solvers, the radial 2D MOC solver and the axial 1D MOC solver, for better accuracy. The expansion was benchmarked with the widely applied C5G7 benchmark models and two fast breeder reactor models, and showed good agreement with the reference Monte Carlo results. In practice, the accurate neutron transport analysis for a full reactor core is still time-consuming and thus limits its application. Therefore, another content of my research is focused on designing a specific hardware based on the reconfigurable computing technique in order to accelerate AGENT computations. It is the first time that the application of this type is used to the reactor physics and neutron transport for reactor design. The most time consuming part of the AGENT algorithm was identified. Moreover, the architecture of the AGENT acceleration system was designed based on the analysis. Through the parallel computation on the specially designed, highly efficient architecture, the acceleration design on FPGA acquires high performance at the much lower working frequency than CPUs. The whole design simulations show that the acceleration design would be able to speedup large scale AGENT computations about 20 times. The high performance AGENT acceleration system will drastically shortening the computation time for 3D full-core neutron transport analysis, making the AGENT methodology unique and advantageous, and thus supplies the possibility to extend the application range of neutron transport analysis in either industry engineering or academic research.

Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

NASA Astrophysics Data System (ADS)

Cisneros, Anselmo Tomas, Jr.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of economic assumptions about the electricity market to evaluate the economic implications of design decisions. The optimal PB-FHR design---Mark 1 PB-FHR---is described along with a detailed summary of its performance characteristics including: the burnup, the burnup evolution, temperature reactivity coefficients, the power distribution, radiation damage distributions, control element worths, decay heat curves and tritium production rates. The Mk1 PB-FHR satisfies the PB-FHR safety criteria. The fuel, moderator (pebble core, pebble shell, graphite matrix, TRISO layers) and coolant have global negative temperature reactivity coefficients and the fuel temperatures are well within their limits.

Development of the University of Washington Biofuels and Biobased Chemicals Process Laboratory

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

Gustafson, Richard

2014-02-04

The funding from this research grant enabled us to design and build a bioconversion steam explosion reactor and ancillary equipment such as a high pressure boiler and a fermenter to support the bioconversion process research. This equipment has been in constant use since its installation in 2012. Following are research projects that it has supported: • Investigation of novel chip production method in biofuels production • Investigation of biomass refining following steam explosion • Several studies on use of different biomass feedstocks • Investigation of biomass moisture content on pretreatment efficacy. • Development of novel instruments for biorefinery process controlmore » Having this equipment was also instrumental in the University of Washington receiving a $40 million grant from the US Department of Agriculture for biofuels development as well as several other smaller grants. The research that is being done with the equipment from this grant will facilitate the establishment of a biofuels industry in the Pacific Northwest and enable the University of Washington to launch a substantial biofuels and bio-based product research program.« less

Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.

2007-01-01

In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is expected within the next 30 to 50 years, as predicted by the Hubbert model and confirmed by other global energy consumption prognoses. Having invested national resources into the development of NGNP, the technology and experience accumulated during the project needs to be documented clearly and in sufficient detail for young engineers coming on-board at both DOE and NASA to acquire it. Hands on training on reactor operation, test rigs of turbomachinery, and heat exchanger components, as well as computational tools will be needed. Senior scientist/engineers involved with the development of NGNP should also be encouraged to participate as lecturers, instructors, or adjunct professors at local universities having engineering (mechanical, electrical, nuclear/chemical, and/or materials) as one of their fields of study.

DEFE0023863 Final Report, Technology for GHG Emission Reduction and CostCompetitive MilSpec Jet Fuel Production using CTL

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

Hartvigsen, Joseph J; Dimick, Paul; Laumb, Jason D

Ceramatec Inc, in collaboration with IntraMicron (IM), the Energy & Environmental Research Center (EERC) and Sustainable Energy Solutions, LLC (SES), have completed a three-year research project integrating their respective proprietary technologies in key areas to demonstrate production of a jet fuel from coal and biomass sources. The project goals and objectives were to demonstrate technology capable of producing a “commercially-viable quantity” of jet fuel and make significant progress toward compliance with Section 526 of the Energy Independence and Security Act of 2007 (EISA 2007 §526) lifecycle greenhouse gas (GHG) emissions requirements. The Ceramatec led team completed the demonstration of nominalmore » 2 bbl/day Fischer-Tropsch (FT) synthesis pilot plant design, capable of producing a nominal 1 bbl/day in the Jet-A/JP-8 fraction. This production rate would have a capacity of 1,000 gallons of jet fuel per month and provide the design basis of a 100 bbl/day module producing over 2,000 gallons of jet fuel per day. Co-gasification of coal-biomass blends enables a reduction of lifecycle greenhouse gas emissions from equivalent conventional petroleum derived fuel basis. Due to limits of biomass availability within an economic transportation range, implementation of a significant biomass feed fraction will require smaller plants than current world scale CTL and GTL FT plants. Hence a down-scaleable design is essential. The pilot plant design leverages Intramicron’s MicroFiber Entrapped Catalyst (MFEC) support which increases the catalyst bed thermal conductivity two orders of magnitude, allowing thermal management of the FT reaction exotherm in much larger reactor tubes. In this project, single tube reactors having 4.5 inch outer diameter and multi-tube reactors having 4 inch outer diameters were operated, with productivities as high as 1.5 gallons per day per linear foot of reactor tube. A significant reduction in tube count results from the use of large diameter reactor tubes, with an associated reduction in reactor cost. The pilot plant was designed with provisions for product collection capable of operating with conventional wax producing FT catalysts but was operated with a Chevron hybrid wax-free FT catalyst. Process simplification enabled by elimination of the wax hydrocracking process unit provides economic advantages in scaling to biomass capable plant sizes. Intramicron also provided a sulfur capture system based on their Oxidative Sulfur Removal (OSR) catalyst process. The integrated sulfur removal and FT systems were operated with syngas produced by the Transport Reactor Development Unit (TRDU) gasifier at the University of North Dakota EERC. SES performed modeling of their cryogenic carbon capture process on the energy, cost and CO2 emissions impact of the Coal-biomass synthetic fuel process.« less

Potential Ways to Address Shortage Situations of 99Mo/99mTc.

PubMed

Filzen, Leah M; Ellingson, Lacey R; Paulsen, Andrew M; Hung, Joseph C

2017-03-01

99m Tc, the most common radioisotope used in nuclear medicine, is produced in a nuclear reactor from the decay of 99 Mo. There are only a few aging nuclear reactors around the world that produce 99 Mo, and one of the major contributors, the National Research Universal (Canada), ceased production on October 31, 2016. The National Research Universal produced approximately 40% of the world's 99 Mo supply, so with its shut down, shortages of 99 Mo/ 99m Tc are expected. Methods: Nuclear pharmacies and nuclear medicine departments throughout the United States were contacted and asked to provide their strategies for coping with a shortage of 99 Mo/ 99m Tc. Each of these strategies was evaluated on the basis of its effectiveness for conserving 99m Tc while still meeting the needs of the patients. Results: From the responses, the following 6 categories of strategies, in order of importance, were compiled: contractual agreements with commercial nuclear pharmacies, alternative imaging protocols, changes in imaging schedules, software use, generator management, and reduction of ordered doses or elimination of backup doses. Conclusion: The supply chain of 99 Mo/ 99m Tc is quite fragile; therefore, being aware of the most appropriate coping strategies is crucial. It is essential to build a strong collaboration between the nuclear pharmacy and nuclear medicine department during a shortage situation. With both nuclear medicine departments and nuclear pharmacies implementing viable strategies, such as the ones proposed, the amount of 99m Tc available during a shortage situation can be maximized. © 2017 by the Society of Nuclear Medicine and Molecular Imaging.

NATCRCTR: One-dimensional thermal-hydraulics analysis code for natural-circulation TRIGA reactors

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

Feltus, M.A.; Rubinaccio, G.

1996-12-31

The Pennsylvania State University nuclear engineering department is evaluating the upgrade of the Reed College (Portland, Oregon) TRIGA reactor from 250 kW to 1 MW in two areas: thermal-hydraulics and steady-state neutronics analysis. This analysis was initiated as a cooperative effort between Penn State and Reed College as a training project for two International Atomic Energy Agency (IAEA) fellows from Ghana. The two Ghanaian IAEA fellows were assisted by G. Rubinaccio, an undergraduate, who undertook the task of writing the new computer programs for the thermal-hydraulic and physics evaluation as a three-credit special design project course. The Reed College TRIGA,more » which has a fixed graphite radial reflector, is cooled by natural circulation, without external cross-flow; whereas, the Penn State Breazeale Reactor has significant crossflow into its sides. To model the Reed TRIGA, the NATCRCTR program has been developed from first principles using the following assumptions: 1. The core is surrounded by the fixed reflector structure, which acts as a one-dimensional channel. 2. The core inlet temperature distribution is constant at the core bottom. 3. The axial heat flux distribution is a chopped cosine shape. 4. The heat transfer in the fuel is primarily in the radial directions. 5. A small gap between the fuel and cladding exists. The NATCRCTR code is used to find the peak centerline fuel, gap, and cladding surface temperatures, based on assumed flux and engineering peaking factors.« less

Neutron spectrum determination in a sub-critical assembly using the multi-disc neutron activation technique

NASA Astrophysics Data System (ADS)

Koseoglou, P.; Vagena, E.; Stoulos, S.; Manolopoulou, M.

2016-09-01

Neutron spectrum of the sub-critical nuclear assembly-reactor of Aristotle University of Thessaloniki was measured at three radial distances from the reactor core. The neutron activation technique was applied irradiating 15 thick foils - disc of various elements at each position. The data of 38 (n, γ), (n, p) and (n, α) reactions were analyzed for specific activity determination. Discs instead of foils were used due to the relevant low neutron flux, so the gamma self-absorption as well as the neutron self-shielding factors has been calculated using GEANT simulations in order to determine the activity induced. The specific activities calculated for all isotopes studied were the input to the SANDII code, which was built specifically for the neutron spectrum de-convolution when the neutron activation technique is used. For the optimization of the results a technique was applied in order to minimize the influence of the initial-"guessed" spectrum shape SANDII uses. The neutron spectrum estimated presents a peak in the regions of (i) thermal neutrons ranged between 0.001 and 1 eV peaking at neutron energy ∼0.1 eV and (ii) fast neutrons ranged between 0.1 and 20 MeV peaking at neutron energy ∼1.2 MeV. The reduction of thermal neutrons is higher than the fast one as the distance from the reactor core increases since thermal neutrons capture by natural U-fuel has higher cross section than the fast neutrons.

Performance Analyses of 38 kWe Turbo-Machine Unit for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

Gallo, Bruno M.; El-Genk, Mohamed S.

2008-01-01

This paper developed a design and investigated the performance of 38 kWe turbo-machine unit for space nuclear reactor power systems with Closed Brayton Cycle (CBC) energy conversion. The compressor and turbine of this unit are scaled versions of the NASA's BRU developed in the sixties and seventies. The performance results of turbo-machine unit are calculated for rotational speed up to 45 krpm, variable reactor thermal power and system pressure, and fixed turbine and compressor inlet temperatures of 1144 K and 400 K. The analyses used a detailed turbo-machine model developed at the University of New Mexico that accounts for the various energy losses in the compressor and turbine and the effect of compressibility of the He-Xe (40 mole/g) working fluid with increased flow rate. The model also accounts for the changes in the physical and transport properties of the working fluid with temperature and pressure. Results show that a unit efficiency of 24.5% is achievable at rotation speed of 45 krpm and system pressure of 0.75 MPa, assuming shaft and electrical generator efficiencies of 86.7% and 90%. The corresponding net electric power output of the unit is 38.5 kWe, the flow rate of the working fluid is 1.667 kg/s, the pressure ratio and polytropic efficiency for the compressor are 1.60 and 83.1%, and 1.51 and 88.3% for the turbine.

Analytical methods in the high conversion reactor core design

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

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

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

Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

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

Liu, Maolong; Ryals, Matthew; Ali, Amir

2016-08-01

A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentallymore » investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.« less

Tungsten - Yttrium Based Nuclear Structural Materials

NASA Astrophysics Data System (ADS)

Ramana, Chintalapalle; Chessa, Jack; Martinenz, Gustavo

2013-04-01

The challenging problem currently facing the nuclear science community in this 21st century is design and development of novel structural materials, which will have an impact on the next-generation nuclear reactors. The materials available at present include reduced activation ferritic/martensitic steels, dispersion strengthened reduced activation ferritic steels, and vanadium- or tungsten-based alloys. These materials exhibit one or more specific problems, which are either intrinsic or caused by reactors. This work is focussed towards tungsten-yttrium (W-Y) based alloys and oxide ceramics, which can be utilized in nuclear applications. The goal is to derive a fundamental scientific understanding of W-Y-based materials. In collaboration with University of Califonia -- Davis, the project is designated to demonstrate the W-Y based alloys, ceramics and composites with enhanced physical, mechanical, thermo-chemical properties and higher radiation resistance. Efforts are focussed on understanding the microstructure, manipulating materials behavior under charged-particle and neutron irradiation, and create a knowledge database of defects, elemental diffusion/segregation, and defect trapping along grain boundaries and interfaces. Preliminary results will be discussed.

[Microbial community in the Anammox process of thermal denitration tail liquid].

PubMed

Li, Jin; Yu, Deshuang; Zhao, Dan; Wang, Xiaochen

2014-12-01

An anaerobic sequencing batch reactor (ASBR) was used to treat thermal denitration tail liquid and microbial community was studied. Activated sludge was taken from the reactor for scanning electron microscope analysis. The images showed that the dominant cells in the flora were oval cocci. Its diameter was about 0.7 μm. Through a series of molecular biology methods such as extracting total DNA from the sludge, PCR amplification, positive clone authentication and sequencing, we obtained the 16S rDNA sequences of the flora. Phylogenetic tree and clone library were established. The universal bacteria primers of 27F-1492R PCR amplification system obtained 85 clones and could be divided into 21 OTUS. The proportions were as follows: Proteobacteria 61.18%; Acidobacteria 17.65%; Chlorobi 8.24%; Chlorofexi 5.88%; Gemmatimonadetes 3.53%; Nitrospirae 2.35% and Planctomycetes 1.18%. The specific anammox bacterial primers of pla46rc-630r and AMX368-AMX820 PCR amplification system obtained 45 clones. They were divided into 3 OTUS. Candidatus brocadia sp. occupied 95.6% and unknown strains occupied 4.4%.

Future U.S. supply of Mo-99 production through fission based LEU/LEU technology.

PubMed

Welsh, James; Bigles, Carmen I; Valderrabano, Alejandro

Coquí RadioPharmaceuticals Corp. (Coquí) has the goal of establishing a medical isotope production facility for securing a continuous domestic supply of the radioisotope molybdenum-99 for U.S. citizens. Coquí will use an LEU/LEU proven and implemented open pool, light-water, 10 MW, reactor design. The facility is being designed with twin reactors for reliability an on-site hot lab chemical processing and a waste conditioning area and a possible generator producing radio-chemistry lab. Coquí identified a 25 acre site adjacent to an existing industrial park in northern central Florida. This land was gifted and transferred to Coquí by the University of Florida Foundation. We are in the process of developing licensing documents related to the facility. The construction permit application for submission to the U.S. Nuclear Regulatory Commission is currently being prepared. Submission is scheduled for mid to late 2015. Community reaction to the proposed development has been positive. We expect to create 220 permanent jobs and we have an anticipated to be operational by 2020.

A Risk Analysis of the Molybdenum-99 Supply Chain Using Bayesian Networks

NASA Astrophysics Data System (ADS)

Liang, Jeffrey Ryan

The production of Molybdenum-99 (99Mo) is critical to the field of nuclear medicine, where it is utilized in roughly 80% of all nuclear imaging procedures. In October of 2016, the National Research Universal (NRU) reactor in Canada, which historically had the highest 99Mo production capability worldwide, ceased routine production and will be permanently shut down in 2018. This loss of capacity has led to widespread concern over the ability of the 99Mo supply chain and to meet demand. There is significant disagreement among analyses from trade groups, governments, and other researchers, predicting everything from no significant impact to major worldwide shortages. Using Bayesian networks, this research focused on modeling the 99Mo supply chain to quantify how a disrupting event, such as the unscheduled downtime of a reactor, will impact the global supply. This not only includes quantifying the probability of a shortage occurring, but also identifying which nodes in the supply chain introduce the most risk to better inform decision makers on where future facilities or other risk mitigation techniques should be applied.

OVERVIEW OF NEUTRON MEASUREMENTS IN JET FUSION DEVICE.

PubMed

Batistoni, P; Villari, R; Obryk, B; Packer, L W; Stamatelatos, I E; Popovichev, S; Colangeli, A; Colling, B; Fonnesu, N; Loreti, S; Klix, A; Klosowski, M; Malik, K; Naish, J; Pillon, M; Vasilopoulou, T; De Felice, P; Pimpinella, M; Quintieri, L

2017-10-05

The design and operation of ITER experimental fusion reactor requires the development of neutron measurement techniques and numerical tools to derive the fusion power and the radiation field in the device and in the surrounding areas. Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case in ITER and power plant studies. The required radiation transport calculations are extremely challenging because of the large physical extent of the reactor plant, the complexity of the geometry, and the combination of deep penetration and streaming paths. This article reports the experimental activities which are carried-out at JET to validate the neutronics measurements methods and numerical tools used in ITER and power plant design. A new deuterium-tritium campaign is proposed in 2019 at JET: the unique 14 MeV neutron yields produced will be exploited as much as possible to validate measurement techniques, codes, procedures and data currently used in ITER design thus reducing the related uncertainties and the associated risks in the machine operation. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

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

A 2-D/1-D transverse leakage approximation based on azimuthal, Fourier moments

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

Stimpson, Shane G.; Collins, Benjamin S.; Downar, Thomas

Here, the MPACT code being developed collaboratively by Oak Ridge National Laboratory and the University of Michigan is the primary deterministic neutron transport solver within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). In MPACT, the two-dimensional (2-D)/one-dimensional (1-D) scheme is the most commonly used method for solving neutron transport-based three-dimensional nuclear reactor core physics problems. Several axial solvers in this scheme assume isotropic transverse leakages, but work with the axial S N solver has extended these leakages to include both polar and azimuthal dependence. However, explicit angular representation can be burdensome for run-time and memory requirements. The workmore » here alleviates this burden by assuming that the azimuthal dependence of the angular flux and transverse leakages are represented by a Fourier series expansion. At the heart of this is a new axial SN solver that takes in a Fourier expanded radial transverse leakage and generates the angular fluxes used to construct the axial transverse leakages used in the 2-D-Method of Characteristics calculations.« less

A 2-D/1-D transverse leakage approximation based on azimuthal, Fourier moments

DOE PAGES

Stimpson, Shane G.; Collins, Benjamin S.; Downar, Thomas

2017-01-12

Here, the MPACT code being developed collaboratively by Oak Ridge National Laboratory and the University of Michigan is the primary deterministic neutron transport solver within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). In MPACT, the two-dimensional (2-D)/one-dimensional (1-D) scheme is the most commonly used method for solving neutron transport-based three-dimensional nuclear reactor core physics problems. Several axial solvers in this scheme assume isotropic transverse leakages, but work with the axial S N solver has extended these leakages to include both polar and azimuthal dependence. However, explicit angular representation can be burdensome for run-time and memory requirements. The workmore » here alleviates this burden by assuming that the azimuthal dependence of the angular flux and transverse leakages are represented by a Fourier series expansion. At the heart of this is a new axial SN solver that takes in a Fourier expanded radial transverse leakage and generates the angular fluxes used to construct the axial transverse leakages used in the 2-D-Method of Characteristics calculations.« less

Monitoring Uranium Transformations Determined by the Evolution of Biogeochemical Processes: Design of Mixed Batch Reactor and Column Studies at Oak Ridge National Laboratory

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

Criddle, Craig S.; Wu, Weimin

2013-04-17

With funds provided by the US DOE, Argonne National Laboratory subcontracted the design of batch and column studies to a Stanford University team with field experience at the ORNL IFRC, Oak Ridge, TN. The contribution of the Stanford group ended in 2011 due to budget reduction in ANL. Over the funded research period, the Stanford research team characterized ORNL IFRC groundwater and sediments and set up microcosm reactors and columns at ANL to ensure that experiments were relevant to field conditions at Oak Ridge. The results of microcosm testing demonstrated that U(VI) in sediments was reduced to U(IV) with themore » addition of ethanol. The reduced products were not uraninite but were instead U(IV) complexes associated with Fe. Fe(III) in solid phase was only partially reduced. The Stanford team communicated with the ANL team members through email and conference calls and face to face at the annual ERSP PI meeting and national meetings.« less

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

NASA Technical Reports Server (NTRS)

VanDyke, Melissa; Martin, James

2004-01-01

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

Fusion Safety Program annual report, fiscal year 1994

NASA Astrophysics Data System (ADS)

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

1995-03-01

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

Development of Innovative Accident Tolerant High Thermal Conductivity UO 2-Diamond Composite Fuel Pellets

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

Tulenko, James; Subhash, Ghatu

2016-01-01

The University of Florida (UF) evaluated a composite fuel consisting of UO 2 powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO 2-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO 2 – SiC and UO 2 – Carbon Nanotube fuel pins. UF ismore » proving with the current research results that the addition of diamond micro particles to UO 2 may greatly enhanced the thermal conductivity of the UO 2 pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.« less

ATRC Neutron Detector Testing Quick Look Report

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

Troy C. Unruh; Benjamin M. Chase; Joy L. Rempe

2013-08-01

As part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program, a joint Idaho State University (ISU) / French Alternative Energies and Atomic Energy Commission (CEA) / Idaho National Laboratory (INL) project was initiated in FY-10 to investigate the feasibility of using neutron sensors to provide online measurements of the neutron flux and fission reaction rate in the ATR Critical Facility (ATRC). A second objective was to provide initial neutron spectrum and flux distribution information for physics modeling and code validation using neutron activation based techniques in ATRC as well as ATR during depressurized operations. Detailed activationmore » spectrometry measurements were made in the flux traps and in selected fuel elements, along with standard fission rate distribution measurements at selected core locations. These measurements provide additional calibration data for the real-time sensors of interest as well as provide benchmark neutronics data that will be useful for the ATR Life Extension Program (LEP) Computational Methods and V&V Upgrade project. As part of this effort, techniques developed by Prof. George Imel will be applied by Idaho State University (ISU) for assessing the performance of various flux detectors to develop detailed procedures for initial and follow-on calibrations of these sensors. In addition to comparing data obtained from each type of detector, calculations will be performed to assess the performance of and reduce uncertainties in flux detection sensors and compare data obtained from these sensors with existing integral methods employed at the ATRC. The neutron detectors required for this project were provided to team participants at no cost. Activation detectors (foils and wires) from an existing, well-characterized INL inventory were employed. Furthermore, as part of an on-going ATR NSUF international cooperation, the CEA sent INL three miniature fission chambers (one for detecting fast flux and two for detecting thermal flux) with associated electronics for assessment. In addition, Prof. Imel, ISU, has access to an inventory of Self-Powered Neutron Detectors (SPNDs) with a range of response times as well as Back-to-Back (BTB) fission chambers from prior research he conducted at the Transient REActor Test Facility (TREAT) facility and Neutron RADiography (NRAD) reactors. Finally, SPNDs from the National Atomic Energy Commission of Argentina (CNEA) were provided in connection with the INL effort to upgrade ATR computational methods and V&V protocols that are underway as part of the ATR LEP. Work during fiscal year 2010 (FY10) focussed on design and construction of Experiment Guide Tubes (EGTs) for positioning the flux detectors in the ATRC N-16 locations as well as obtaining ATRC staff concurrence for the detector evaluations. Initial evaluations with CEA researchers were also started in FY10 but were cut short due to reactor reliability issues. Reactor availability issues caused experimental work to be delayed during FY11/12. In FY13, work resumed; and evaluations were completed. The objective of this "Quick Look" report is to summarize experimental activities performed from April 4, 2013 through May 16, 2013.« less

(High temperature flaw assessment procedure)

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

Ruggles, M.B.

1990-06-01

The Electric Power Research Institute (EPRI), the Japanese Central Research Institute of Electric Power Industry (CRIEPI), and the British Nuclear Electric (NE) are conducting joint studies in the field of liquid metal reactor development. The traveler is currently responsible for the EPRI/CRIEPI/NE High-Temperature Flaw Assessment Procedure activities at the Oak Ridge National Laboratory (ORNL). The traveler participated, on behalf of EPRI, in the EPRI/CRIEPI/NE specialist working session, the purpose of which was to produce the interim High-Temperature Flaw Assessment guide. The traveler also led discussions on the High-Temperature Flaw Assessment Procedure Phase 2 program plan, and on the plan formore » a new joint EPRI/CRIEPI/NE study in Inelastic Behavior and Failure Criteria for Modified 9Cr--1Mo Steel. The traveler visited Profs. K. Ikegami, Y. Asada, N. Ohno, T. Inoue, and K. Kaneko at the Tokyo Institute of Technology, the University of Tokyo, Nagoya University, Kyoto University, and Science University of Tokyo, respectively to hold discussions on research advances in the areas of high-temperature fracture mechanics, inelastic material behavior, and constitutive modeling. In addition, the traveler visited Kajima Corp. and Ohbayashi Corp. Technical Research Institute to collect information on research in the area of fiber reinforced concrete.« less

Advanced applications of cosmic-ray muon radiography

NASA Astrophysics Data System (ADS)

Perry, John

The passage of cosmic-ray muons through matter is dominated by the Coulomb interaction with electrons and atomic nuclei. The muon's interaction with electrons leads to continuous energy loss and stopping through the process of ionization. The muon's interaction with nuclei leads to angular diffusion. If a muon stops in matter, other processes unfold, as discussed in more detail below. These interactions provide the basis for advanced applications of cosmic-ray muon radiography discussed here, specifically: 1) imaging a nuclear reactor with near horizontal muons, and 2) identifying materials through the analysis of radiation lengths weighted by density and secondary signals that are induced by cosmic-ray muon trajectories. We have imaged a nuclear reactor, type AGN-201m, at the University of New Mexico, using data measured with a particle tracker built from a set of sealed drift tubes, the Mini Muon Tracker (MMT). Geant4 simulations were compared to the data for verification and validation. In both the data and simulation, we can identify regions of interest in the reactor including the core, moderator, and shield. This study reinforces our claims for using muon tomography to image reactors following an accident. Warhead and special nuclear materials (SNM) imaging is an important thrust for treaty verification and national security purposes. The differentiation of SNM from other materials, such as iron and aluminum, is useful for these applications. Several techniques were developed for material identification using cosmic-ray muons. These techniques include: 1) identifying the radiation length weighted by density of an object and 2) measuring the signals that can indicate the presence of fission and chain reactions. By combining the radiographic images created by tracking muons through a target plane with the additional fission neutron and gamma signature, we are able to locate regions that are fissionable from a single side. The following materials were imaged with this technique: aluminum, concrete, steel, lead, and uranium. Provided that there is sufficient mass, U-235 could be differentiated from U-238 through muon induced fission.

γ{\\hbox-}Flash suppression using a gated photomultiplier assembled with an LaBr3(Ce) detector to measure fast neutron capture reactions

NASA Astrophysics Data System (ADS)

Hara, K. Y.; Harada, H.; Toh, Y.; Hori, J.

2013-09-01

A gated photomultiplier tube (PMT) assembled with an LaBr3(Ce) detector was applied toward the prompt γ{hbox-}ray measurement of fast neutron capture reactions. Time-of-flight measurements of the neutron capture reactions of Cl and Al were performed using the 46-MeV electron linear accelerator at the Kyoto University Research Reactor Institute (KURRI) as a pulsed neutron source. The photomultiplier gating technique effectively suppressed the saturation of the PMT output and extended the energy region of the TOF measurement.

Evaluation of Neutron Response of Criticality Accident Alarm System Detector to Quasi-Monoenergetic 24 keV Neutrons

NASA Astrophysics Data System (ADS)

Tsujimura, Norio; Yoshida, Tadayoshi; Yashima, Hiroshi

The criticality accident alarm system (CAAS), which was recently developed and installed at the Japan Atomic Energy Agency's Tokai Reprocessing Plant, consists of a plastic scintillator combined with a cadmium-lined polyethylene moderator and thereby responds to both neutrons and gamma rays. To evaluate the neutron absorbed dose rate response of the CAAS detector, a 24 keV quasi-monoenergetic neutron irradiation experiment was performed at the B-1 facility of the Kyoto University Research Reactor. The detector's evaluated neutron response was confirmed to agree reasonably well with prior computer-predicted responses.

Tritium safety study using Caisson Assembly (CATS) at TPL/JAEA

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

Hayashi, T.; Kobayashi, K.; Iwai, Y.

Tritium confinement is required as the most important safety Junction for a fusion reactor. In order to demonstrate the confinement performance experimentally, an unique equipment, called CATS: Caisson Assembly for Tritium Safety study, was installed in Tritium Process Laboratory of Japan Atomic Energy Agency and operated for about 10 years. Tritium confinement and migration data in CATS have been accumulated and dynamic simulation code was accumulated using these data. Contamination and decontamination behavior on various materials and new safety equipment functions have been investigated under collaborations with a lot of laboratories and universities. (authors)

Visualization of cavitation phenomena in a Diesel engine fuel injection nozzle by neutron radiography

NASA Astrophysics Data System (ADS)

Takenaka, N.; Kadowaki, T.; Kawabata, Y.; Lim, I. C.; Sim, C. M.

2005-04-01

Visualization of cavitation phenomena in a Diesel engine fuel injection nozzle was carried out by using neutron radiography system at KUR in Research Reactor Institute in Kyoto University and at HANARO in Korea Atomic Energy Research Institute. A neutron chopper was synchronized to the engine rotation for high shutter speed exposures. A multi-exposure method was applied to obtain a clear image as an ensemble average of the synchronized images. Some images were successfully obtained and suggested new understanding of the cavitation phenomena in a Diesel engine fuel injection nozzle.

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

Wieser, Patti; Hopkins, David

The DOE Princeton Plasma Physics Laboratory (PPPL) collaborates to develop fusion as a safe, clean and abundant energy source for the future. This video discusses PPPL's research and development on plasma, the fourth state of matter. In this simulation of plasma turbulence inside PPPL's National Spherical Torus Experiment, the colorful strings represent higher and lower electron density in turbulent plasma as it circles around a donut-shaped fusion reactor; red and orange are higher density. This image is among those featured in the slide show, "Plasmas are Hot and Fusion is Cool," a production of PPPL and the Princeton University Broadcastmore » Center.« less

The Preignition and Autoignition Oxidation of Alternatives to Petroleum Derived JP-8 and their Surrogate Components in a Pressurized Flow Reactor and Single Cylinder Research Engine

DTIC Science & Technology

2009-09-01

sample probe consisted of TIG welding the 3/8” sample probe shaft to the sample probe tip (Koert, 1990 and Lenhert, 2004b). Silver solder was...was performed in the Drexel University Machine Shop. Conventional TIG welding was sufficient for welding the 3/8” O.D. tube to the sample probe tip...However, to TIG weld the thermocouple and the glass lined tube to the sample probe tip, extreme care had to be taken so as not to damage the

Attrition Resistant Iron-Based Catalysts For F-T SBCRs

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

Adeyinka A. Adeyiga

2006-01-31

The Fischer-Tropsch (F-T) reaction provides a way of converting coal-derived synthesis gas (CO+ H{sub 2}) to liquid fuels. Since the reaction is highly exothermic, one of the major problems in control of the reaction is heat removal. Recent work has shown that the use of slurry bubble column reactors (SBCRs) can largely solve this problem. The use of iron-(FE) based catalysts is attractive not only due to their low cost and ready availability, but also due to their high water-gas shift activity which makes it possible to use these catalysts with low H{sub 2}/CO ratios. However, a serious problem withmore » the use of Fe catalysts in a SBCR is their tendency to undergo attrition. This can cause fouling/plugging of downstream filters and equipment; makes the separation of catalyst from the oil/wax product very difficult, if not impossible; and results in a steady loss of catalyst from the reactor. Under a previous Department of Energy (DOE)/University Research Grant (UCR) grant, Hampton University reported, for the first time, the development of demonstrably attrition-resistant Fe F-T synthesis catalysts having good activity, selectivity, and attrition resistance. These catalysts were prepared by spray drying Fe catalysts with potassium (K), copper (Cu), and silica (SiO{sub 2}) as promoters. SiO{sub 2} was also used as a binder for spray drying. These catalysts were tested for activity and selectivity in a laboratory-scale fixed-bed reactor. Fundamental understanding of attrition is being addressed by incorporating suitable binders into the catalyst recipe. This has resulted in the preparation of a spray dried HPR-43 catalyst having average particle size (aps) of 70 {micro}m with high attrition resistance. This HPR-43 attrition resistant, active and selective catalyst gave 95% CO conversion through 125 hours of testing in a fixed-bed at 270 C, 1.48 MPa, H{sub 2}/CO=0.67 and 2.0 NL/g-cat/h with C{sub 5+} selectivity of >78% and methane selectivity of less than 5% at an {alpha} of 0.9. Research is proposed to enable further development and optimization of these catalysts by (1) better understanding the role and interrelationship of various catalyst composition and preparation parameters on attrition resistance, activity, and selectivity of these catalysts, (2) the presence of sulfide ions on a precipitated iron catalyst, and (3) the effect of water on sulfided iron F-T catalysts for its activity, selectivity, and attrition. Catalyst preparations will be based on spray drying. The research employed, among other measurements, attrition testing and F-T synthesis at high pressure. Catalyst activity and selectivity is evaluated using a small fixed-bed reactor and a continuous stirred tank reactor (CSTR). The catalysts were prepared by co-precipitation, followed by binder addition and spray drying at 250 C in a 1-m-diameter, 2-m-tall spray dryer. The binder silica content was varied from 0 to 20 wt%. The results show that the use of small amounts of precipitated SiO{sub 2} alone in spray-dried Fe catalysts can result in good attrition resistance. All catalysts investigated with SiO2 wt% {le} 12 produced fines less than 10 wt% during the jet cup attrition test, making them suitable for long-term use in a slurry bubble column reactor. Thus, concentration rather than the type of SiO{sub 2} incorporated into catalyst has a more critical impact on catalyst attrition resistance of spray-dried Fe catalysts. Lower amounts of SiO{sub 2} added to a catalyst give higher particle densities and therefore higher attrition resistances. In order to produce a suitable SBCR catalyst, however, the amount of SiO{sub 2} added has to be optimized to provide adequate surface area, particle density, and attrition resistance. Two of the catalysts with precipitated and binder silica were tested in Texas A&M University's CSTR (Autoclave Engineers). The two catalysts were also tested at The Center for Applied Energy Research in Lexington, Kentucky of the University of Kentucky. Spray-dried catalysts with compositions 100 Fe/5 Cu/4.2 K/11 (P) SiO{sub 2} and 100 Fe/5 Cu/4.2 K/1.1 (B) SiO{sub 2} have excellent selectivity characteristics (low methane and high C{sub 5+} yields), but their productivity and stability (deactivation rate) need to be improved. Mechanical integrity (attrition strength) of these two catalysts was markedly dependent upon their morphological features. The attrition strength of the catalyst made out of largely spherical particles (1.1 (B) SiO{sub 2}) was considerably higher than that of the catalyst consisting of irregularly shaped particles (11 (P) SiO{sub 2}).« less

Development of Ultra-Fine Multigroup Cross Section Library of the AMPX/SCALE Code Packages

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

Jeon, Byoung Kyu; Sik Yang, Won; Kim, Kang Seog

The Consortium for Advanced Simulation of Light Water Reactors Virtual Environment for Reactor Applications (VERA) neutronic simulator MPACT is being developed by Oak Ridge National Laboratory and the University of Michigan for various reactor applications. The MPACT and simplified MPACT 51- and 252-group cross section libraries have been developed for the MPACT neutron transport calculations by using the AMPX and Standardized Computer Analyses for Licensing Evaluations (SCALE) code packages developed at Oak Ridge National Laboratory. It has been noted that the conventional AMPX/SCALE procedure has limited applications for fast-spectrum systems such as boiling water reactor (BWR) fuels with very highmore » void fractions and fast reactor fuels because of its poor accuracy in unresolved and fast energy regions. This lack of accuracy can introduce additional error sources to MPACT calculations, which is already limited by the Bondarenko approach for resolved resonance self-shielding calculation. To enhance the prediction accuracy of MPACT for fast-spectrum reactor analyses, the accuracy of the AMPX/SCALE code packages should be improved first. The purpose of this study is to identify the major problems of the AMPX/SCALE procedure in generating fast-spectrum cross sections and to devise ways to improve the accuracy. For this, various benchmark problems including a typical pressurized water reactor fuel, BWR fuels with various void fractions, and several fast reactor fuels were analyzed using the AMPX 252-group libraries. Isotopic reaction rates were determined by SCALE multigroup (MG) calculations and compared with continuous energy (CE) Monte Carlo calculation results. This reaction rate analysis revealed three main contributors to the observed differences in reactivity and reaction rates: (1) the limitation of the Bondarenko approach in coarse energy group structure, (2) the normalization issue of probability tables, and (3) neglect of the self-shielding effect of resonance-like cross sections at high energy range such as (n,p) cross section of Cl35. The first error source can be eliminated by an ultra-fine group (UFG) structure in which the broad scattering resonances of intermediate-weight nuclides can be represented accurately by a piecewise constant function. A UFG AMPX library was generated with modified probability tables and tested against various benchmark problems. The reactivity and reaction rates determined with the new UFG AMPX library agreed very well with respect to Monte Carlo Neutral Particle (MCNP) results. To enhance the lattice calculation accuracy without significantly increasing the computational time, performing the UFG lattice calculation in two steps was proposed. In the first step, a UFG slowing-down calculation is performed for the corresponding homogenized composition, and UFG cross sections are collapsed into an intermediate group structure. In the second step, the lattice calculation is performed for the intermediate group level using the condensed group cross sections. A preliminary test showed that the condensed library reproduces the results obtained with the UFG cross section library. This result suggests that the proposed two-step lattice calculation approach is a promising option to enhance the applicability of the AMPX/SCALE system to fast system analysis.« less

JPRS Report, Science & Technology, China: Energy.

DTIC Science & Technology

1992-03-30

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

Walter Greiner: In Memoriam

NASA Astrophysics Data System (ADS)

Zen Vasconcellos, César; Coelho, Helio T.; Hess, Peter Otto

Walter Greiner (29 October 1935 - 6 October 2016) was a German theoretical physicist. His scientific research interests include the thematic areas of atomic physics, heavy ion physics, nuclear physics, elementary particle physics (particularly quantum electrodynamics and quantum chromodynamics). He is most known in Germany for his series of books in theoretical physics, but he is also well known around the world. Greiner was born on October 29, 1935, in Neuenbau, Sonnenberg, Germany. He studied physics at the University of Frankfurt (Goethe University in Frankfurt Am Main), receiving in this institution a BSci in physics and a Master’s degree in 1960 with a thesis on plasma-reactors, and a PhD in 1961 at the University of Freiburg under Hans Marshal, with a thesis on the nuclear polarization in μ-mesic atoms. During the period of 1962 to 1964 he was assistant professor at the University of Maryland, followed by a position as research associate at the University of Freiburg, in 1964. Starting in 1965, he became a full professor at the Institute for Theoretical Physics at Goethe University until 2003. Greiner has been a visiting professor to many universities and laboratories, including Florida State University, the University of Virginia, the University of California, the University of Melbourne, Vanderbilt University, Yale University, Oak Ridge National Laboratory and Los Alamos National Laboratory. In 2003, with Wolf Singer, he was the founding Director of the Frankfurt Institute for Advanced Studies (FIAS), and gave lectures and seminars in elementary particle physics. He died on October 6, 2016 at the age of 80. Walter Greiner was an excellent teacher, researcher, friend. And he was a great supporter of the series of events known by the acronyms IWARA - International Workshop on Astronomy and Relativistic Astrophysics, STARS - Caribbean Symposium on Cosmology, Gravitation, Nuclear and Astroparticle Physics, and SMFNS - International Symposium on Strong Electromagnetic Fields and Neutron Stars. Walter Greiner left us. But his memory will remain always alive among us who have had the privilege of knowing him and enjoy his wisdom and joy of living.

Application of a fuzzy logic control system for continuous anaerobic digestion of low buffered, acidic energy crops as mono-substrate.

PubMed

Scherer, P; Lehmann, K; Schmidt, O; Demirel, B

2009-02-15

A fuzzy logic control (FLC) system was developed at the Hamburg University of Applied Sciences (HAW Hamburg) for operation of biogas reactors running on energy crops. Three commercially available measuring parameters, namely pH, the methane (CH4) content, and the specific gas production rate (spec. GPR = m(3)/kg VS/day) were included. The objective was to avoid stabilization of pH with use of buffering supplements, like lime or manure. The developed FLC system can cover most of all applications, such as a careful start-up process and a gentle recovery strategy after a severe reactor failure, also enabling a process with a high organic loading rate (OLR) and a low hydraulic retention time (HRT), that is, a high throughput anaerobic digestion process with a stable pH and CH4 content. A precondition for a high load process was the concept of interval feeding, for example, with 8 h of interval. The FLC system was proved to be reliable during the long term fermentation studies over 3 years in one-stage, completely stirred tank reactors (CSTR) with acidic beet silage as mono-input (pH 3.3-3.4). During fermentation of the fodder beet silage (FBS), a stable HRT of 6.0 days with an OLR of up to 15 kg VS/m(3)/day and a volumetric GPR of 9 m(3)/m(3)/day could be reached. The FLC enabled an automatic recovery of the digester after two induced severe reactor failures. In another attempt to prove the feasibility of the FLC, substrate FBS was changed to sugar beet silage (SBS), which had a substantially lower buffering capacity than that of the FBS. With SBS, the FLC accomplished a stable fermentation at a pH level between 6.5 and 6.6, and a volatile fatty acid level (VFA) below 500 mg/L, but the FLC had to interact and to change the substrate dosage permanently. In a further experiment, the reactor temperature was increased from 41 to 50 degrees C. Concomitantly, the specific GPR, pH and CH4 dropped down. Finally, the FLC automatically enabled a complete recovery in 16 days.

Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

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

A. J. Palmer; DC Haggard; J. W. Herter

High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type Nmore » thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina insulation and molybdenum sheath. The most current version of the High Temperature Irradiation Resistant Thermocouple (HTIR-TC) based on molybdenum/niobium alloys, and developed at Idaho National Laboratory, was also tested.« less

Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

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

Palmer, A. J.; Haggard, DC; Herter, J. W.

High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to bemore » only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard-fired alumina insulation and a molybdenum sheath. The most current version of the High Temperature Irradiation Resistant Thermocouple, based on molybdenum/niobium alloys and developed at Idaho National Laboratory, was also tested. (authors)« less

Biofilm reactors for industrial bioconversion processes: employing potential of enhanced reaction rates

PubMed Central

Qureshi, Nasib; Annous, Bassam A; Ezeji, Thaddeus C; Karcher, Patrick; Maddox, Ian S

2005-01-01

This article describes the use of biofilm reactors for the production of various chemicals by fermentation and wastewater treatment. Biofilm formation is a natural process where microbial cells attach to the support (adsorbent) or form flocs/aggregates (also called granules) without use of chemicals and form thick layers of cells known as "biofilms." As a result of biofilm formation, cell densities in the reactor increase and cell concentrations as high as 74 gL-1 can be achieved. The reactor configurations can be as simple as a batch reactor, continuous stirred tank reactor (CSTR), packed bed reactor (PBR), fluidized bed reactor (FBR), airlift reactor (ALR), upflow anaerobic sludge blanket (UASB) reactor, or any other suitable configuration. In UASB granular biofilm particles are used. This article demonstrates that reactor productivities in these reactors have been superior to any other reactor types. This article describes production of ethanol, butanol, lactic acid, acetic acid/vinegar, succinic acid, and fumaric acid in addition to wastewater treatment in the biofilm reactors. As the title suggests, biofilm reactors have high potential to be employed in biotechnology/bioconversion industry for viable economic reasons. In this article, various reactor types have been compared for the above bioconversion processes. PMID:16122390

Hybrid activated sludge/biofilm process for the treatment of municipal wastewater in a cold climate region: a case study.

PubMed

Di Trapani, Daniele; Christensso, Magnus; Odegaard, Hallvard

2011-01-01

A hybrid activated sludge/biofilm process was investigated for wastewater treatment in a cold climate region. This process, which contains both suspended biomass and biofilm, usually referred as IFAS process, is created by introducing plastic elements as biofilm carrier media into a conventional activated sludge reactor. In the present study, a hybrid process, composed of an activated sludge and a moving bed biofilm reactor was used. The aim of this paper has been to investigate the performances of a hybrid process, and in particular to gain insight the nitrification process, when operated at relatively low MLSS SRT and low temperatures. The results of a pilot-scale study carried out at the Department of Hydraulic and Environmental Engineering at the Norwegian University of Science and Technology in Trondheim are presented. The experimental campaign was divided into two periods. The pilot plant was first operated with a constant HRT of 4.5 hours, while in the second period the influent flow was increased so that HRT was 3.5 hours. The average temperature was near 11.5°C in the overall experimental campaign. The average mixed liquor SRT was 5.7 days. Batch tests on both carriers and suspended biomass were performed in order to evaluate the nitrification rate of the two different biomasses. The results demonstrated that this kind of reactor can efficiently be used for the upgrading of conventional activated sludge plant for achieving year-round nitrification, also in presence of low temperatures, and without the need of additional volumes.

Automating High-Precision X-Ray and Neutron Imaging Applications with Robotics

DOE PAGES

Hashem, Joseph Anthony; Pryor, Mitch; Landsberger, Sheldon; ...

2017-03-28

Los Alamos National Laboratory and the University of Texas at Austin recently implemented a robotically controlled nondestructive testing (NDT) system for X-ray and neutron imaging. This system is intended to address the need for accurate measurements for a variety of parts and, be able to track measurement geometry at every imaging location, and is designed for high-throughput applications. This system was deployed in a beam port at a nuclear research reactor and in an operational inspection X-ray bay. The nuclear research reactor system consisted of a precision industrial seven-axis robot, 1.1-MW TRIGA research reactor, and a scintillator-mirror-camera-based imaging system. Themore » X-ray bay system incorporated the same robot, a 225-keV microfocus X-ray source, and a custom flat panel digital detector. The robotic positioning arm is programmable and allows imaging in multiple configurations, including planar, cylindrical, as well as other user defined geometries that provide enhanced engineering evaluation capability. The imaging acquisition device is coupled with the robot for automated image acquisition. The robot can achieve target positional repeatability within 17 μm in the 3-D space. Flexible automation with nondestructive imaging saves costs, reduces dosage, adds imaging techniques, and achieves better quality results in less time. Specifics regarding the robotic system and imaging acquisition and evaluation processes are presented. In conclusion, this paper reviews the comprehensive testing and system evaluation to affirm the feasibility of robotic NDT, presents the system configuration, and reviews results for both X-ray and neutron radiography imaging applications.« less

VRA Modeling, phase 1

NASA Technical Reports Server (NTRS)

Kindt, Louis M.; Mullins, Michael E.; Hand, David W.; Kline, Andrew A.

1995-01-01

The destruction of organic contaminants in waste water for closed systems, such as that of Space Station, is crucial due to the need for recycling the waste water. A co-current upflow bubble column using oxygen as the gas phase oxidant and packed with catalyst particles consisting of a noble metal on an alumina substrate is being developed for this process. The objective of this study is to develop a plug-flow model that will predict the performance of this three phase reactor system in destroying a multicomponent mixture of organic contaminants in water. Mass balances on a series of contaminants and oxygen in both the liquid and gas phases are used to develop this model. These mass balances incorporate the gas-to-liquid and liquid-to-particle mass transfer coefficients, the catalyst effectiveness factor, and intrinsic reaction rate. To validate this model, a bench scale reactor has been tested at Michigan Technological University at elevated pressures (50-83 psig,) and a temperature range of 200 to 290 F. Feeds consisting of five dilute solutions of ethanol (approx. 10 ppm), chlorobenzene (approx. 20 ppb), formaldehyde (approx. 100 ppb), dimethyl sulfoxide (DMSO approx. 300 ppb), and urea (approx. 20 ppm) in water were tested individually with an oxygen mass flow rate of 0.009 lb/h. The results from these individual tests were used to develop the kinetic parameter inputs necessary for the computer model. The computer simulated results are compared to the experimental data obtained for all 5 components run in a mixture on the differential test column for a range of reactor contact times.

Experimental Investigations in a Reactor Cavity Cooling System with Advanced Instrumentation for the Study of Instabilities, Oscillations, and Transients

NASA Astrophysics Data System (ADS)

Tompkins, Casey A.

A research team at University of Wisconsin - Madison designed and constructed a 1/4 height scaled experimental facility to study two-phase natural circulation cooling in a water-based reactor cavity cooling system (WRCCS) for decay heat removal in an advanced high temperature reactor. The facility is capable of natural circulation operation scaled for simulated decay heat removal (up to 28.5 kW m-2 (45 kW) input power, which is equivalent to 14.25 kW m-2 (6.8 MW) at full scale) and pressurized up to 2 bar. The UW-WRCCS facility has been used to study instabilities and oscillations observed during natural circulation flow due to evaporation of the water inventory. During two-phase operation, the system exhibits flow oscillations and excursions, which cause thermal oscillations in the structure. This can cause degradation in the mechanical structure at welds and limit heat transfer to the coolant. The facility is equipped with wire mesh sensors (WMS) that enable high-resolution measurements of the void fraction and steam velocities in order to study the instability's and oscillation's growth and decay during transient operation. Multiple perturbations to the system's operating point in pressure and inlet throttling have shown that the oscillatory behavior present under normal two-phase operating conditions can be damped and removed. Furthermore, with steady-state modeling it was discovered that a flow regime transition instability is the primary cause of oscillations in the UW-WRCCS facility under unperturbed conditions and that proper orifice selection can move the system into a stable operating regime.

Automating High-Precision X-Ray and Neutron Imaging Applications with Robotics

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

Hashem, Joseph Anthony; Pryor, Mitch; Landsberger, Sheldon

Los Alamos National Laboratory and the University of Texas at Austin recently implemented a robotically controlled nondestructive testing (NDT) system for X-ray and neutron imaging. This system is intended to address the need for accurate measurements for a variety of parts and, be able to track measurement geometry at every imaging location, and is designed for high-throughput applications. This system was deployed in a beam port at a nuclear research reactor and in an operational inspection X-ray bay. The nuclear research reactor system consisted of a precision industrial seven-axis robot, 1.1-MW TRIGA research reactor, and a scintillator-mirror-camera-based imaging system. Themore » X-ray bay system incorporated the same robot, a 225-keV microfocus X-ray source, and a custom flat panel digital detector. The robotic positioning arm is programmable and allows imaging in multiple configurations, including planar, cylindrical, as well as other user defined geometries that provide enhanced engineering evaluation capability. The imaging acquisition device is coupled with the robot for automated image acquisition. The robot can achieve target positional repeatability within 17 μm in the 3-D space. Flexible automation with nondestructive imaging saves costs, reduces dosage, adds imaging techniques, and achieves better quality results in less time. Specifics regarding the robotic system and imaging acquisition and evaluation processes are presented. In conclusion, this paper reviews the comprehensive testing and system evaluation to affirm the feasibility of robotic NDT, presents the system configuration, and reviews results for both X-ray and neutron radiography imaging applications.« less

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014

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

Soelberg, Renae

2014-11-01

Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014 Highlights Rory Kennedy and Sarah Robertson attended the American Nuclear Society Winter Meeting and Nuclear Technology Expo in Anaheim, California, Nov. 10-13. ATR NSUF exhibited at the technology expo where hundreds of meeting participants had an opportunity to learn more about ATR NSUF. Dr. Kennedy briefed the Nuclear Engineering Department Heads Organization (NEDHO) on the workings of the ATR NSUF. • Rory Kennedy, James Cole and Dan Ogden participated in a reactor instrumentation discussion with Jean-Francois Villard and Christopher Destouches of CEA and several members of themore » INL staff. • ATR NSUF received approval from the NE-20 office to start planning the annual Users Meeting. The meeting will be held at INL, June 22-25. • Mike Worley, director of the Office of Innovative Nuclear Research (NE-42), visited INL Nov. 4-5. Milestones Completed • Recommendations for the Summer Rapid Turnaround Experiment awards were submitted to DOE-HQ Nov. 12 (Level 2 milestone due Nov. 30). Major Accomplishments/Activities • The University of California, Santa Barbara 2 experiment was unloaded from the GE-2000 at HFEF. The experiment specimen packs will be removed and shipped to ORNL for PIE. • The Terrani experiment, one of three FY 2014 new awards, was completed utilizing the Advanced Photon Source MRCAT beamline. The experiment investigated the chemical state of Ag and Pd in SiC shell of irradiated TRISO particles via X-ray Absorption Fine Structure (XAFS) spectroscopy. Upcoming Meetings/Events • The ATR NSUF program review meeting will be held Dec. 9-10 at L’Enfant Plaza. In addition to NSUF staff and users, NE-4, NE-5 and NE-7 representatives will attend the meeting. Awarded Research Projects Boise State University Rapid Turnaround Experiments (14-485 and 14-486) Nanoindentation and TEM work on the T91, HT9, HCM12A and 9Cr ODS specimens has been completed at CAES by Boise State PI Janelle Wharry and Cory Dolph. PI Corey Dolph returned in early November to complete their research by performing nanoindentation on unirradiated specimens that will be used as a baseline for their research.« 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.

Thermal mechanical analysis of applications with internal heat generation

NASA Astrophysics Data System (ADS)

Govindarajan, Srisharan Garg

The radioactive tracer Technetium-99m is widely used in medical imaging and is derived from its parent isotope Molybedenum-99 (Mo-99) by radioactive decay. The majority of Molybdenum-99 (Mo-99) produced internationally is extracted from high enriched uranium (HEU) dispersion targets that have been irradiated. To alleviate proliferation risks associated with HEU-based targets, the use of non-HEU sources is being mandated. However, the conversion of HEU to LEU based dispersion targets affects the Mo-99 available for chemical extraction. A possible approach to increase the uranium density, to recover the loss in Mo-99 production-per-target, is to use an LEU metal foil placed within an aluminum cladding to form a composite structure. The target is expected to contain the fission products and to dissipate the generated heat to the reactor coolant. In the event of interfacial separation, an increase in the thermal resistance could lead to an unacceptable rise in the LEU temperature and stresses in the target. The target can be deemed structurally safe as long as the thermally induced stresses are within the yield strength of the cladding and welds. As with the thermal and structural safety of the annular target, the thermally induced deflection of the BORALRTM-based control blades, used by the University of Missouri Research Reactor (MURRRTM ), during reactor operation has been analyzed. The boron, which is the neutron absorber in BORAL, and aluminum mixture (BORAL meat) and the aluminum cladding are bonded together through powder metallurgy to establish an adherent bonded plate. As the BORAL absorbs both neutron particles and gamma rays, there is volumetric heat generation and a corresponding rise in temperature. Since the BORAL meat and aluminum cladding materials have different thermal expansion coefficients, the blade may have a tendency to deform as the blade temperature changes and the materials expand at different rates. In addition to the composite nature of the control blade, spatial variations in temperature within the control blade occur from the non-uniform heat generation within the BORAL as a result of the non-uniform thermal neutron flux along the longitudinal direction when the control blade is partially withdrawn. There is also variation in the heating profile through the thickness and about the circumferential width of the control blade. Mathematical curve-fits are generated for the non-uniform volumetric heat generation profile caused by the thermal neutron absorption and the functions are applied as heating conditions within a finite element model of the control blade built using the commercial finite element code Abaqus FEA. The finite element model is solved as a fully coupled thermal mechanical problem as in the case of the annular target. The resulting deflection is compared with the channel gap to determine if there is a significant risk of the control blade binding during reactor operation. Hence, this dissertation will consist of two sections. The first section will seek to present the thermal and structural safety analyses of the annular targets for the production of molybdenum-99. Since there hasn't been any detailed, documented, study on these annular targets in the past, the work complied in this dissertation will help to understand the thermal-mechanical behavior and failure margins of the target during in-vessel irradiation. As the work presented in this dissertation provides a general performance analysis envelope for the annular target, the tools developed in the process can also be used as useful references for future analyses that are specific to any reactor. The numerical analysis approach adopted and the analytical models developed, can also be applied to other applications, outside the Mo-99 project domain, where internal heat generation exists such as in electronic components and nuclear reactor control blades. The second section will focus on estimating the thermally induced deflection and hence establish operational safety of the BORAL control blades used at the Missouri University Research Reactor (MURR) to support their relicensing efforts with the Nuclear Regulatory Commission (NRC). The common theme in both these sections is the nuclear heat source, high heat flux, non-uniform heating, composite structures and differential thermal expansion. The goal is to establish the target and component operational safety, and also provide documented analysis that can be referred to in the future.

Stabilization of sulfuric acid dimers by ammonia, methylamine, dimethylamine, and trimethylamine

NASA Astrophysics Data System (ADS)

Jen, Coty N.; McMurry, Peter H.; Hanson, David R.

2014-06-01

This study experimentally explores how ammonia (NH3), methylamine (MA), dimethylamine (DMA), and trimethylamine (TMA) affect the chemical formation mechanisms of electrically neutral clusters that contain two sulfuric acid molecules (dimers). Dimers may also contain undetectable compounds, such as water or bases, that evaporate upon ionization and sampling. Measurements were conducted using a glass flow reactor which contained a steady flow of humidified nitrogen with sulfuric acid concentrations of 107 to 109 cm-3. A known molar flow rate of a basic gas was injected into the flow reactor. The University of Minnesota Cluster Chemical Ionization Mass Spectrometer was used to measure the resulting sulfuric acid vapor and cluster concentrations. It was found that, for a given concentration of sulfuric acid vapor, the dimer concentration increases with increasing concentration of the basic gas, eventually reaching a plateau. The base concentrations at which the dimer concentrations saturate suggest NH3 < MA < TMA ≲ DMA in forming stabilized sulfuric acid dimers. Two heuristic models for cluster formation by acid-base reactions are developed to interpret the data. The models provide ranges of evaporation rate constants that are consistent with observations and leads to an analytic expression for nucleation rates that is consistent with atmospheric observations.

Assessment of RANS and LES Turbulence Modeling for Buoyancy-Aided/Opposed Forced and Mixed Convection

NASA Astrophysics Data System (ADS)

Clifford, Corey; Kimber, Mark

2017-11-01

Over the last 30 years, an industry-wide shift within the nuclear community has led to increased utilization of computational fluid dynamics (CFD) to supplement nuclear reactor safety analyses. One such area that is of particular interest to the nuclear community, specifically to those performing loss-of-flow accident (LOFA) analyses for next-generation very-high temperature reactors (VHTR), is the capacity of current computational models to predict heat transfer across a wide range of buoyancy conditions. In the present investigation, a critical evaluation of Reynolds-averaged Navier-Stokes (RANS) and large-eddy simulation (LES) turbulence modeling techniques is conducted based on CFD validation data collected from the Rotatable Buoyancy Tunnel (RoBuT) at Utah State University. Four different experimental flow conditions are investigated: (1) buoyancy-aided forced convection; (2) buoyancy-opposed forced convection; (3) buoyancy-aided mixed convection; (4) buoyancy-opposed mixed convection. Overall, good agreement is found for both forced convection-dominated scenarios, but an overly-diffusive prediction of the normal Reynolds stress is observed for the RANS-based turbulence models. Low-Reynolds number RANS models perform adequately for mixed convection, while higher-order RANS approaches underestimate the influence of buoyancy on the production of turbulence.

Manufacturing Technology of Ceramic Pebbles for Breeding Blanket.

PubMed

Lo Frano, Rosa; Puccini, Monica; Stefanelli, Eleonora; Del Serra, Daniele; Malquori, Stefano

2018-05-02

An open issue for the fusion power reactor is the choice of breeding blanket material. The possible use of Helium-Cooled Pebble Breeder ceramic material in the form of pebble beds is of great interest worldwide as demonstrated by the numerous studies and research on this subject. Lithium orthosilicate (Li₄SiO₄) is a promising breeding material investigated in this present study because the neutron capture of Li-6 allows the production of tritium, 6Li (n, t) 4He. Furthermore, lithium orthosilicate has the advantages of low activation characteristics, low thermal expansion coefficient, high thermal conductivity, high density and stability. Even if they are far from the industrial standard, a variety of industrial processes have been proposed for making orthosilicate pebbles with diameters of 0.1⁻1 mm. However, some manufacturing problems have been observed, such as in the chemical stability (agglomeration phenomena). The aim of this study is to provide a new methodology for the production of pebbles based on the drip casting method, which was jointly developed by the DICI-University of Pisa and Industrie Bitossi. Using this new (and alternative) manufacturing technology, in the field of fusion reactors, appropriately sized ceramic pebbles could be produced for use as tritium breeders.

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

Hanan, N. A.; Matos, J. E.

At The request of the Czech Technical University in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. For core configurations C1 and C2, criticality calculations were done for cases with all control rodsmore » at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were done for the C1 core configuration. Finally the reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations.« less

Development of a Molecular Tagging Velocimetry Technique for Non-Intrusive Velocity Measurements in Low-Speed Gas Flows

NASA Technical Reports Server (NTRS)

Andre, M. A.; Bardet, P. M.; Cadell, S. R.; Woods, B.; Burns, R. A.; Danehy, P. M.

2017-01-01

N2O molecular tagging velocimetry (N2O-MTV) is developed for use in very-high-temperature reactor environments. Tests were carried out to determine the optimum excitation wavelength, tracer concentration, and timing parameters for the laser system. Using NO tracers obtained from photo-dissociation of N2O, velocity profiles are successfully obtained in air, nitrogen, and helium for a large range of parameters: temperature from 295 to 781 K, pressure from 1 to 3 bars, with a velocity precision of 0.01 m/s. Furthermore, by using two read pulses at adjustable time delays, the velocity dynamic range can be increased. An unprecedented dynamic range of 5,000 has been obtained to successfully resolve the flow during a helium blowdown from 1000 m/s down to 0.2 m/s. This technique is also applied to the high-temperature test facility (HTTF) at Oregon State University (OSU) during a depressurized condition cooldown (DCC) event. Details of these measurements are presented in a companion paper. This technique shows a strong potential for fundamental understanding of gas flows in nuclear reactors and to provide benchmark experimental data to validate numerical simulations.

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

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

Kelley, D.B.

1995-12-31

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

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2011 CFR

2011-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

Nuclear Reactors. Revised.

ERIC Educational Resources Information Center

Hogerton, John F.

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

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2012 CFR

2012-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

Advanced reactors and novel reactions for the conversion of triglyceride based oils into high quality renewable transportation fuels

NASA Astrophysics Data System (ADS)

Linnen, Michael James

Sustainable energy continues to grow more important to all societies, leading to the research and development of a variety of alternative and renewable energy technologies. Of these, renewable liquid transportation fuels may be the most visible to consumers, and this visibility is further magnified by the long-term trend of increasingly expensive petroleum fuels that the public consumes. While first-generation biofuels such as biodiesel and fuel ethanol have been integrated into the existing fuel infrastructures of several countries, the chemical differences between them and their petroleum counterparts reduce their effectiveness. This gives rise to the development and commercialization of second generation biofuels, many of which are intended to have equivalent properties to those of their petroleum counterparts. In this dissertation, the primary reactions for a second-generation biofuel process, known herein as the University of North Dakota noncatalytic cracking process (NCP), have been studied at the fundamental level and improved. The NCP is capable of producing renewable fuels and chemicals that are virtually the same as their petroleum counterparts in performance and quality (i.e., petroleum-equivalent). In addition, a novel analytical method, FIMSDIST was developed which, within certain limitations, can increase the elution capabilities of GC analysis and decrease sample processing times compared to other high resolution methods. These advances are particularly useful for studies of highly heterogeneous fuel and/or organic chemical intermediates, such as those studied for the NCP. However the data from FIMSDIST must be supplemented with data from other methods such as for certain carboxylic acid, to provide accurate, comprehensive results, From a series of TAG cracking experiments that were performed, it was found that coke formation during cracking is most likely the result of excessive temperature and/or residence time in a cracking reactor. Based on this, a tubular cracking reactor was developed that could operate continuously without coke formation. The design also was proven to be scalable. Yields from the reactor were determined under a variety of conditions in order to predict the outputs from the NCP and to establish relationships/correlations between operating parameters and the product distribution. These studies led to the conclusion that the most severe operating conditions which do not induce coking are optimal over the experimental domain. In order to develop economical deoxygenation catalysts for use within the NCP, a series of experiments were performed using nickel catalysts, demonstrating that nickel catalysts could outperform their predecessor, a high cost palladium-based catalyst. A nickel catalyst was then tested in a packed bed reactor in order to determine suitable operating conditions for its commercial utilization in packed bed reactors.

Nuclear reactor construction with bottom supported reactor vessel

DOEpatents

Sharbaugh, John E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

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

D. Kokkinos

2005-04-28

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

Heat transfer analysis of cylindrical anaerobic reactors with different sizes: a heat transfer model.

PubMed

Liu, Jiawei; Zhou, Xingqiu; Wu, Jiangdong; Gao, Wen; Qian, Xu

2017-10-01

The temperature is the essential factor that influences the efficiency of anaerobic reactors. During the operation of the anaerobic reactor, the fluctuations of ambient temperature can cause a change in the internal temperature of the reactor. Therefore, insulation and heating measures are often used to maintain anaerobic reactor's internal temperature. In this paper, a simplified heat transfer model was developed to study heat transfer between cylindrical anaerobic reactors and their surroundings. Three cylindrical reactors of different sizes were studied, and the internal relations between ambient temperature, thickness of insulation, and temperature fluctuations of the reactors were obtained at different reactor sizes. The model was calibrated by a sensitivity analysis, and the calibrated model was well able to predict reactor temperature. The Nash-Sutcliffe model efficiency coefficient was used to assess the predictive power of heat transfer models. The Nash coefficients of the three reactors were 0.76, 0.60, and 0.45, respectively. The model can provide reference for the thermal insulation design of cylindrical anaerobic reactors.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, Robert M.

1983-01-01

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, R.M.

1983-11-08

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

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

Winger, Jeff Allen

Beta-decay spectroscopy provides important information on nuclear structure and properties needed to understand topics as widely varied as fundamental nuclear astrophysics to applied nuclear reactor design. However, there are significant limitations of our knowledge due to an inability to experimentally measure everything. Therefore, it is often necessary to rely on theoretical calculations which need to be vetted with experimental results. The focus of this report will be results from experimental research performed by the Principal Investigator (PI) and his research group at Mississippi State University in which the group played the lead role in proposing, implementing, performing and analyzing themore » experiment. This research was carried out at both the National Superconduction Cyclotron Laboratory (NSCL) at Michigan State University and the Holifield Radioactive Ion Beam Facility (HRIBF) at Oak Ridge National Laboratory. The primary emphasis of the research was the use of \\bdec spectroscopy as a tool to understand the evolution of nuclear structure in neutron-rich nuclei which could then be applied to improve theory and to increase the overall knowledge of nuclear structure.« less

A study of the evaporation of heterogeneous water droplets under active heating

NASA Astrophysics Data System (ADS)

Piskunov, Maxim; Legros, Jean Claude; Strizhak, Pavel

2016-11-01

Using high-speed video registration tools with a sample rate of 102-104 frames per second (fps), we studied the patterns in the evaporation of water droplets containing 1 and 2 mm individual metallic inclusions in a high-temperature gas environment. The materials of choice for the inclusions were steels (AISI 1080 carbon steel and AISI type 316L stainless steel) and pure nickel. We established the lifetimes τh of the liquid droplets under study with a controlled increase in the gas environment temperature up to 900 K. We also considered the physical aspects behind the τh distribution in the experiments conducted and specified the conditions for more effective cooling of metallic inclusions. Following the experimental research findings, a method was devised for effective reactor vessel cooling to avoid a meltdown at a nuclear power plant. The optimization of heat and mass transfer modes was performed within the framework of the strategic plan for the development of National Research Tomsk Polytechnic University as one of the world-leading universities.

Early warning indicators for monitoring the process failure of anaerobic digestion system of food waste.

PubMed

Li, Lei; He, Qingming; Wei, Yunmei; He, Qin; Peng, Xuya

2014-11-01

To determine reliable state parameters which could be used as early warning indicators of process failure due to the acidification of anaerobic digestion of food waste, three mesophilic anaerobic digesters of food waste with different operation conditions were investigated. Such parameters as gas production, methane content, pH, concentrations of volatile fatty acid (VFA), alkalinity and their combined indicators were evaluated. Results revealed that operation conditions significantly affect the responses of parameters and thus the optimal early warning indicators of each reactor differ from each other. None of the single indicators was universally valid for all the systems. The universally valid indicators should combine several parameters to supply complementary information. A combination of total VFA, the ratio of VFA to total alkalinity (VFA/TA) and the ratio of bicarbonate alkalinity to total alkalinity (BA/TA) can reflect the metabolism of the digesting system and realize rapid and effective early warning. Copyright © 2014 Elsevier Ltd. All rights reserved.

Laboratory Astrophysics Prize: Laboratory Astrophysics with Nuclei

NASA Astrophysics Data System (ADS)

Wiescher, Michael

2018-06-01

Nuclear astrophysics is concerned with nuclear reaction and decay processes from the Big Bang to the present star generation controlling the chemical evolution of our universe. Such nuclear reactions maintain stellar life, determine stellar evolution, and finally drive stellar explosion in the circle of stellar life. Laboratory nuclear astrophysics seeks to simulate and understand the underlying processes using a broad portfolio of nuclear instrumentation, from reactor to accelerator from stable to radioactive beams to map the broad spectrum of nucleosynthesis processes. This talk focuses on only two aspects of the broad field, the need of deep underground accelerator facilities in cosmic ray free environments in order to understand the nucleosynthesis in stars, and the need for high intensity radioactive beam facilities to recreate the conditions found in stellar explosions. Both concepts represent the two main frontiers of the field, which are being pursued in the US with the CASPAR accelerator at the Sanford Underground Research Facility in South Dakota and the FRIB facility at Michigan State University.

A unified model of quarks and leptons with a universal texture zero

NASA Astrophysics Data System (ADS)

de Medeiros Varzielas, Ivo; Ross, Graham G.; Talbert, Jim

2018-03-01

We show that a universal texture zero in the (1,1) position of all fermionic mass matrices, including heavy right-handed Majorana neutrinos driving a type-I see-saw mechanism, can lead to a viable spectrum of mass, mixing and CP violation for both quarks and leptons, including (but not limited to) three important postdictions: the Cabibbo angle, the charged lepton masses, and the leptonic `reactor' angle. We model this texture zero with a non-Abelian discrete family symmetry that can easily be embedded in a grand unified framework, and discuss the details of the phenomenology after electroweak and family symmetry breaking. We provide an explicit numerical fit to the available data and obtain excellent agreement with the 18 observables in the charged fermion and neutrino sectors with just 9 free parameters. We further show that the vacua of our new scalar familon fields are readily aligned along desired directions in family space, and also demonstrate discrete gauge anomaly freedom at the relevant scale of our effective theory.

University Engagement at INL

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

Morrell, Sean Robert; Rynes, Amanda Renee

2014-07-01

There are currently over 900 facilities in over 170 countries which fall under International Atomic Energy Agency (IAEA) safeguards. As additional nations look to purse civilian nuclear programs or to expand infrastructure already in place, the number of reactors and accompanying facilities as well as the quantity of material has greatly increased. Due to the breadth of the threat and the burden placed on the IAEA as nuclear applications expand, it has become increasingly important that safeguards professionals have a strong understanding of both the technical and political aspects of nonproliferation starting early in their career. To begin overcoming thismore » challenge, Idaho National Laboratory, has partnered with local universities to deliver a graduate level nuclear engineering course that covers both aspects of the field with a focus on safeguards applications. To date over 60 students across multiple disciplines have participated in this course with many deciding to transition into a nonproliferation area of focus in both their academic and professional careers.« less

Nuclear reactor neutron shielding

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

Speaker, Daniel P; Neeley, Gary W; Inman, James B

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactormore » cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.« less

Reactor pressure vessel head vents and methods of using the same

DOEpatents

Gels, John L; Keck, David J; Deaver, Gerald A

2014-10-28

Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

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

10 CFR 52.167 - Issuance of manufacturing license.

Code of Federal Regulations, 2010 CFR

2010-01-01

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

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

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

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

Process and apparatus for adding and removing particles from pressurized reactors

DOEpatents

Milligan, John D.

1983-01-01

A method for adding and removing fine particles from a pressurized reactor is provided, which comprises connecting the reactor to a container, sealing the container from the reactor, filling the container with particles and a liquid material compatible with the reactants, pressurizing the container to substantially the reactor pressure, removing the seal between the reactor and the container, permitting particles to fall into or out of the reactor, and resealing the container from the reactor. An apparatus for adding and removing particles is also disclosed.

Effects of imperfect mixing on low-density polyethylene reactor dynamics

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

Villa, C.M.; Dihora, J.O.; Ray, W.H.

1998-07-01

Earlier work considered the effect of feed conditions and controller configuration on the runaway behavior of LDPE autoclave reactors assuming a perfectly mixed reactor. This study provides additional insight on the dynamics of such reactors by using an imperfectly mixed reactor model and bifurcation analysis to show the changes in the stability region when there is imperfect macroscale mixing. The presence of imperfect mixing substantially increases the range of stable operation of the reactor and makes the process much easier to control than for a perfectly mixed reactor. The results of model analysis and simulations are used to identify somemore » of the conditions that lead to unstable reactor behavior and to suggest ways to avoid reactor runaway or reactor extinction during grade transitions and other process operation disturbances.« less

Startup of reactors for anoxic ammonium oxidation: experiences from the first full-scale anammox reactor in Rotterdam.

PubMed

van der Star, Wouter R L; Abma, Wiebe R; Blommers, Dennis; Mulder, Jan-Willem; Tokutomi, Takaaki; Strous, Marc; Picioreanu, Cristian; van Loosdrecht, Mark C M

2007-10-01

The first full-scale anammox reactor in the world was started in Rotterdam (NL). The reactor was scaled-up directly from laboratory-scale to full-scale and treats up to 750 kg-N/d. In the initial phase of the startup, anammox conversions could not be identified by traditional methods, but quantitative PCR proved to be a reliable indicator for growth of the anammox population, indicating an anammox doubling time of 10-12 days. The experience gained during this first startup in combination with the availability of seed sludge from this reactor, will lead to a faster startup of anammox reactors in the future. The anammox reactor type employed in Rotterdam was compared to other reactor types for the anammox process. Reactors with a high specific surface area like the granular sludge reactor employed in Rotterdam provide the highest volumetric loading rates. Mass transfer of nitrite into the biofilm is limiting the conversion of those reactor types that have a lower specific surface area. Now the first full-scale commercial anammox reactor is in operation, a consistent and descriptive nomenclature is suggested for reactors in which the anammox process is employed.

A small, 1400 K, reactor for Brayton space power systems.

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.

Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

NASA Technical Reports Server (NTRS)

Roth, R. J.

1976-01-01

The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

KINETICS OF TREAT USED AS A TEST REACTOR

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

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

1962-05-01

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

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

Nuclear reactor having a polyhedral primary shield and removable vessel insulation

DOEpatents

Ekeroth, Douglas E.; Orr, Richard

1993-01-01

A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel.

Nuclear reactor having a polyhedral primary shield and removable vessel insulation

DOEpatents

Ekeroth, D.E.; Orr, R.

1993-12-07

A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel. 5 figures.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, C.W.

1985-02-19

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, Charles W.

1987-01-01

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Summary of NR Program Prometheus Efforts

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

J Ashcroft; C Eshelman

2006-02-08

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

Results and Analysis of the Infrastructure Request for Information (DE-SOL-0008318)

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

Heidrich, Brenden John

2015-07-01

The Department of Energy (DOE) Office of Nuclear Energy (NE) released a request for information (RFI) (DE-SOL-0008318) for “University, National Laboratory, Industry and International Input on Potential Office of Nuclear Energy Infrastructure Investments” on April 13, 2015. DOE-NE solicited information on five specific types of capabilities as well as any others suggested by the community. The RFI proposal period closed on June 19, 2015. From the 26 responses, 34 individual proposals were extracted. Eighteen were associated with a DOE national laboratory, including Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Idaho National Laboratory (INL), Los Alamos National Laboratory (LANL), Pacificmore » Northwest National Laboratory (PNNL) and Sandia National Laboratory (SNL). Oak Ridge National Laboratory (ORNL) was referenced in a proposal as a proposed capability location, although the proposal did not originate with ORNL. Five US universities submitted proposals (Massachusetts Institute of Technology, Pennsylvania State University, Rensselaer Polytechnic Institute, University of Houston and the University of Michigan). Three industrial/commercial institutions submitted proposals (AREVA NP, Babcock and Wilcox (B&W) and the Electric Power Research Institute (EPRI)). Eight major themes emerged from the submissions as areas needing additional capability or support for existing capabilities. Two submissions supported multiple areas. The major themes are: Advanced Manufacturing (AM), High Performance Computing (HPC), Ion Irradiation with X-Ray Diagnostics (IIX), Ion Irradiation with TEM Visualization (IIT), Radiochemistry Laboratories (RCL), Test Reactors, Neutron Sources and Critical Facilities (RX) , Sample Preparation and Post-Irradiation Examination (PIE) and Thermal-Hydraulics Test Facilities (THF).« less

METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

DOEpatents

Reed, G.A.

1961-01-10

A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

Control Means for Reactor

DOEpatents

Manley, J. H.

1961-06-27

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2014 CFR

2014-01-01

... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will inform the... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will accept for... New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they...

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will inform the... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will accept for... New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they...

Contractor's STTR Phase I Final Report- Experimental Analysis and Model Development of Pyrolysis/Combustion of Coal/Biomass in a Bench Scale Spouted Bed Reactor

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

Baird, Benjamin; Loebick, Codruta; Roychoudhury, Subir

During Phase I both experimental evaluation and computational validation of an advanced Spouted Bed Reactor (SBR) approach for biomass and coal combustion was completed. All Phase I objectives were met and some exceeded. Comprehensive insight on SBR operation was achieved via design, fabrication, and testing of a small demonstration unit with pulverized coal and biomass as feedstock at University of Connecticut (UCONN). A scale-up and optimization tool for the next generation of coal and biomass co-firing for reducing GHG emissions was also developed. The predictive model was implemented with DOE’s MFIX computational model and was observed to accurately mimic evenmore » unsteady behavior. An updated Spouted Bed Reactor was fabricated, based on model feedback, and experimentally displayed near ideal behavior. This predictive capability based upon first principles and experimental correlation allows realistic simulation of mixed fuel combustion in these newly proposed power boiler designs. Compared to a conventional fluidized bed the SBR facilitates good mixing of coal and biomass, with relative insensitivity to particle size and densities, resulting in improved combustion efficiency. Experimental data with mixed coal and biomass fuels demonstrated complete oxidation at temperatures as low as 500ºC. This avoids NOx formation and residual carbon in the waste ash. Operation at stoichiometric conditions without requiring cooling or sintering of the carrier was also observed. Oxygen-blown operation were tested and indicated good performance. This highlighted the possibility of operating the SBR at a wide range of conditions suitable for power generation and partial oxidation byproducts. It also supports the possibility of implementing chemical looping (for readily capturing CO 2 and SO x).« less

A chemical and fluid dynamic study of the chemical vapor deposition of aluminum nitride in a vertical reactor

NASA Astrophysics Data System (ADS)

Bather, Wayne Anthony

The metalorganic chemical vapor deposition (MOCVD) growth of compound semiconductors has become important in producing many high performance electronic and optoelectronic devices from the wide bandgaps III-V nitrides, for example, aluminum nitride (AlN). A systematic theoretical and experimental investigation of the chemistry and mass transport process in a MOCVD system can yield predictive models of the deposition process. The chemistries and fluid dynamics of the MOCVD growth of AlN in a vertical reactor is analyzed and characterized in order to parameterize and model the deposition process. A Fourier Transform Infrared (FTIR) spectroscopic study of the predeposition reactions between trimethylaluminum (TMAl) and ammonia (NHsb3) is carried out in a static gas cell to examine the primary homogeneous gas phase reactions, pyrolysis of the reactants, and adduct formation, possibly accompanied by elimination reactions. A series of reactions, based on laboratory studies and literature review, is then proposed to model the deposition process. All pertinent kinetic, thermochemical, and transport properties were obtained. Utilizing a mass transport model, we performed computational fluid dynamics calculations using the FLUENT software package. We determined temperature, velocity, and concentration profiles, along with deposition rates inside the experimental vertical CVD reactor in the Howard University Material Science Research Center of Excellence. Experimental deposition rate data were found to be in good agreement with those predicted from the simulations, thus validating the proposed model. The control of the homogeneous gas phase reaction leading to the formation and subsequent decomposition of the adduct is critical to the formation of device-grade AlN films. Many basic processes occurring during MOCVD of AlN are still not completely understood, and none of the detailed surface reaction mechanisms are known.

Dosimetric and microdosimetric analyses for blood exposed to reactor-derived thermal neutrons.

PubMed

Ali, F; Atanackovic, J; Boyer, C; Festarini, A; Kildea, J; Paterson, L C; Rogge, R; Stuart, M; Richardson, R B

2018-06-06

Thermal neutrons are found in reactor, radiotherapy, aircraft, and space environments. The purpose of this study was to characterise the dosimetry and microdosimetry of thermal neutron exposures, using three simulation codes, as a precursor to quantitative radiobiological studies using blood samples. An irradiation line was designed employing a pyrolytic graphite crystal or-alternatively-a super mirror to expose blood samples to thermal neutrons from the National Research Universal reactor to determine radiobiological parameters. The crystal was used when assessing the relative biological effectiveness for dicentric chromosome aberrations, and other biomarkers, in lymphocytes over a low absorbed dose range of 1.2-14 mGy. Higher exposures using a super mirror will allow the additional quantification of mitochondrial responses. The physical size of the thermal neutron fields and their respective wavelength distribution was determined using the McStas Monte Carlo code. Spinning the blood samples produced a spatially uniform absorbed dose as determined from Monte Carlo N-Particle version 6 simulations. The major part (71%) of the total absorbed dose to blood was determined to be from the 14 N(n,p) 14 C reaction and the remainder from the 1 H(n,γ) 2 H reaction. Previous radiobiological experiments at Canadian Nuclear Laboratories involving thermal neutron irradiation of blood yielded a relative biological effectiveness of 26 ± 7. Using the Particle and Heavy Ion Transport Code System, a similar value of ∼19 for the quality factor of thermal neutrons initiating the 14 N(n,p) 14 C reaction in soft tissue was determined by microdosimetric simulations. This calculated quality factor is of similar high value to the experimentally-derived relative biological effectiveness, and indicates the potential of thermal neutrons to induce deleterious health effects in superficial organs such as cataracts of the eye lens.

An investigation of reactivity effect due to inadvertent filling of the irradiation channels with water in NIRR-1 Nigeria Research Reactor-1.

PubMed

Iliyasu, U; Ibrahim, Y V; Umar, Sadiq; Agbo, S A; Jibrin, Y

2017-05-01

Investigation of reactivity variation due to flooding of the irradiation channels of Nigeria Research Reactor (NIRR-1) a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria using the MCNP code for High Enrich Uranium (HEU) and Low Enrich Uranium (LEU) core has been simulated in this present study. In this work, the excess reactivity worth of flooding HEU core for 1 inner, 2 inner, 3 inner, 4 inner and all inner are 0.318mk, 0.577mk, 0.318mk, 1.204mk and 1.503mk respectively, and outer irradiation channels are 0.119mk, 0.169mk, 0.348mk, 0.438mk and 0.418mk respectively, the highest excess reactivity result from flooding both inner and outer irradiation channels is 2.04mk (±1.72×10 -7 ), the excess reactivity for LEU core was 0.299mk, 0.568mk, 0.896mk, 1.195mk and 1.524mk in the inner irradiation channels, and the outer irradiation channels are 0.129mk, 0.189mk, 0.219mk, 0.269mk and 0.548mk where the highest excess reactivity was 1.942mk (±1.64×10 -7 ) resulting from flooding inner and outer irradiation channels. The reactivity induced by flooding of the irradiation channels of NIRR-1 with water is within design safety limit enshrined in Safety Analysis Report of NIRR-1. The results also compare well with literature. Copyright © 2017 Elsevier Ltd. All rights reserved.

Experimental Analysis and Model Development of Pyrolysis/Combustion of Coal/Biomass in a Bench Scale Spouted Bed Reactor.

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

Baird, Benjamin; Loebick, Codruta; Roychoudhury, Subir

During Phase I both experimental evaluation and computational validation of an advanced Spouted Bed Reactor (SBR) approach for biomass and coal combustion was completed. All Phase I objectives were met and some exceeded. Comprehensive insight on SBR operation was achieved via design, fabrication, and testing of a small demonstration unit with pulverized coal and biomass as feedstock at University of Connecticut (UCONN). A scale-up and optimization tool for the next generation of coal and biomass co-firing for reducing GHG emissions was also developed. The predictive model was implemented with DOE’s MFIX computational model and was observed to accurately mimic evenmore » unsteady behavior. An updated Spouted Bed Reactor was fabricated, based on model feedback, and experimentally displayed near ideal behavior. This predictive capability based upon first principles and experimental correlation allows realistic simulation of mixed fuel combustion in these newly proposed power boiler designs. Compared to a conventional fluidized bed the SBR facilitates good mixing of coal and biomass, with relative insensitivity to particle size and densities, resulting in improved combustion efficiency. Experimental data with mixed coal and biomass fuels demonstrated complete oxidation at temperatures as low as 500C. This avoids NOx formation and residual carbon in the waste ash. Operation at stoichiometric conditions without requiring cooling or sintering of the carrier was also observed. Oxygen-blown operation were tested and indicated good performance. This highlighted the possibility of operating the SBR at a wide range of conditions suitable for power generation and partial oxidation byproducts. It also supports the possibility of implementing chemical looping (for readily capturing CO2 and SOx).« less

Irradiation Testing of Ultrasonic Transducers

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

Daw, Joshua; Tittmann, Bernhard; Reinhardt, Brian

2014-07-30

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of single, small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other efforts include an ultrasonic technique to detect morphologymore » changes (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of existing knowledge of ultrasonic transducer material survivability under irradiation conditions. For this reason, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate promising magnetostrictive and piezoelectric transducer performance in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2 (E> 0.1 MeV). The goal of this research is to characterize magnetostrictive and piezoelectric transducer survivability during irradiation, enabling the development of novel radiation tolerant ultrasonic sensors for use in Material and Test Reactors (MTRs). As such, this test will be an instrumented lead test and real-time transducer performance data will be collected along with temperature and neutron and gamma flux data. The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers.« less

High pressure/high temperature thermogravimetric apparatus. Final report

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

Calo, J.M.; Suuberg, E.M.

1999-12-01

The purpose of this instrumentation grant was to acquire a state-of-the-art, high pressure, high temperature thermogravimetric apparatus (HP/HT TGA) system for the study of the interactions between gases and carbonaceous solids for the purpose of solving problems related to coal utilization and applications of carbon materials. The instrument that we identified for this purpose was manufactured by DMT (Deutsche Montan Technologies)--Institute of Cokemaking and Coal Chemistry of Essen, Germany. Particular features of note include: Two reactors: a standard TGA reactor, capable of 1100 C at 100 bar; and a high temperature (HT) reactor, capable of operation at 1600 C andmore » 100 bar; A steam generator capable of generating steam to 100 bar; Flow controllers and gas mixing system for up to three reaction gases, plus a separate circuit for steam, and another for purge gas; and An automated software system for data acquisition and control. The HP/TP DMT-TGA apparatus was purchased in 1996 and installed and commissioned during the summer of 1996. The apparatus was located in Room 128 of the Prince Engineering Building at Brown University. A hydrogen alarm and vent system were added for safety considerations. The system has been interfaced to an Ametek quadruple mass spectrometer (MA 100), pumped by a Varian V250 turbomolecular pump, as provided for in the original proposed. With this capability, a number of gas phase species of interest can be monitored in a near-simultaneous fashion. The MS can be used in a few different modes. During high pressure, steady-state gasification experiments, it is used to sample, measure, and monitor the reactant/product gases. It can also be used to monitor gas phase species during nonisothermal temperature programmed reaction (TPR) or temperature programmed desorption (TPD) experiments.« less

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

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

Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated inmore » the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.« less

Low temperature pre-treatment of domestic sewage in an anaerobic hybrid or an anaerobic filter reactor.

PubMed

Elmitwalli, Tarek A; Sklyar, Vladimir; Zeeman, Grietje; Lettinga, Gatze

2002-05-01

The pre-treatment of domestic sewage for removal of suspended solids (SS) at a process temperature of 13 degrees C and an hydraulic retention time (HRT) of 4 h was investigated in an anaerobic filter (AF) and anaerobic hybrid (AH) reactor. The AF and the top of the AH reactor consisted of vertical sheets of reticulated polyurethane foam (RPF) with knobs. All biomass in the AF was only in attached form to avoid clogging and sludge washout. The AF reactor showed a significantly higher removal of total and suspended chemical oxygen demand (COD) than the AH reactor, respectively, 55% and 82% in the AF reactor and 34% and 53% in the AH reactor. Because the reactors were operated at a short HRT and low temperature, the hydrolysis, acidification and methanogenesis based on the influent COD were limited to, respectively, 12%, 21% and 23% for the AF reactor and 12%, 17% and 16% for the AH reactor. The excess sludge from the AH reactor was more stabilised and had a better settling capacity and dewaterability. However, the excess sludge from both the AH and AF reactors needed stabilisation. Therefore, the AF reactor is recommended for the pretreatment of domestic sewage at low temperatures.

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

NASA Astrophysics Data System (ADS)

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

2010-11-01

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

NEET Micro-Pocket Fission Detector. Final Project report

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

Unruh, T.; Rempe, Joy; McGregor, Douglas

2014-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Alternative Energies and Atomic Energy Commission, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), is funded by the Nuclear Energy Enabling Technologies (NEET) program to develop and test Micro-Pocket Fission Detectors (MPFDs), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package. When deployed, these sensors will significantly advance flux detection capabilities for irradiation tests in US Material Test Reactors (MTRs). Ultimately, evaluations may lead to a more compact, more accurate, andmore » longer lifetime flux sensor for critical mock-ups, and high performance reactors, allowing several Department of Energy Office of Nuclear Energy (DOE-NE) programs to obtain higher accuracy/higher resolution data from irradiation tests of candidate new fuels and materials. Specifically, deployment of MPFDs will address several challenges faced in irradiations performed at MTRs: Current fission chamber technologies do not offer the ability to measure fast flux, thermal flux and temperature within a single compact probe; MPFDs offer this option. MPFD construction is very different than current fission chamber construction; the use of high temperature materials allow MPFDs to be specifically tailored to survive harsh conditions encountered in-core of high performance MTRs. The higher accuracy, high fidelity data available from the compact MPFD will significantly enhance efforts to validate new high-fidelity reactor physics codes and new multi-scale, multi-physics codes. MPFDs can be built with variable sensitivities to survive the lifetime of an experiment or fuel assembly in some MTRs, allowing for more efficient and cost effective power monitoring. The small size of the MPFDs allows multiple sensors to be deployed, offering the potential to accurately measure the flux and temperature profiles in the reactor. This report summarizes the status at the end of year two of this three year project. As documented in this report, all planned accomplishments for developing this unique new, compact, multipurpose sensor have been completed.« less

Experimental investigations of heat transfer and temperature fields in models simulating fuel assemblies used in the core of a nuclear reactor with a liquid heavy-metal coolant

NASA Astrophysics Data System (ADS)

Belyaev, I. A.; Genin, L. G.; Krylov, S. G.; Novikov, A. O.; Razuvanov, N. G.; Sviridov, V. G.

2015-09-01

The aim of this experimental investigation is to obtain information on the temperature fields and heat transfer coefficients during flow of liquid-metal coolant in models simulating an elementary cell in the core of a liquid heavy metal cooled fast-neutron reactor. Two design versions for spacing fuel rods in the reactor core were considered. In the first version, the fuel rods were spaced apart from one another using helical wire wound on the fuel rod external surface, and in the second version spacer grids were used for the same purpose. The experiments were carried out on the mercury loop available at the Moscow Power Engineering Institute National Research University's Chair of Engineering Thermal Physics. Two experimental sections simulating an elementary cell for each of the fuel rod spacing versions were fabricated. The temperature fields were investigated using a dedicated hinged probe that allows temperature to be measured at any point of the studied channel cross section. The heat-transfer coefficients were determined using the wall temperature values obtained at the moment when the probe thermocouple tail end touched the channel wall. Such method of determining the wall temperature makes it possible to alleviate errors that are unavoidable in case of measuring the wall temperature using thermocouples placed in slots milled in the wall. In carrying out the experiments, an automated system of scientific research was applied, which allows a large body of data to be obtained within a short period of time. The experimental investigations in the first test section were carried out at Re = 8700, and in the second one, at five values of Reynolds number. Information about temperature fields was obtained by statistically processing the array of sampled probe thermocouple indications at 300 points in the experimental channel cross section. Reach material has been obtained for verifying the codes used for calculating velocity and temperature fields in channels with an intricately shaped cross section simulating the flow pass sections for liquid-metal coolants cooling the core of nuclear reactors.

Establishment and assessment of code scaling capability

NASA Astrophysics Data System (ADS)

Lim, Jaehyok

In this thesis, a method for using RELAP5/MOD3.3 (Patch03) code models is described to establish and assess the code scaling capability and to corroborate the scaling methodology that has been used in the design of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. It was sponsored by the United States Nuclear Regulatory Commission (USNRC) under the program "PUMA ESBWR Tests". PUMA-E facility was built for the USNRC to obtain data on the performance of the passive safety systems of the General Electric (GE) Nuclear Energy Economic Simplified Boiling Water Reactor (ESBWR). Similarities between the prototype plant and the scaled-down test facility were investigated for a Gravity-Driven Cooling System (GDCS) Drain Line Break (GDLB). This thesis presents the results of the GDLB test, i.e., the GDLB test with one Isolation Condenser System (ICS) unit disabled. The test is a hypothetical multi-failure small break loss of coolant (SB LOCA) accident scenario in the ESBWR. The test results indicated that the blow-down phase, Automatic Depressurization System (ADS) actuation, and GDCS injection processes occurred as expected. The GDCS as an emergency core cooling system provided adequate supply of water to keep the Reactor Pressure Vessel (RPV) coolant level well above the Top of Active Fuel (TAF) during the entire GDLB transient. The long-term cooling phase, which is governed by the Passive Containment Cooling System (PCCS) condensation, kept the reactor containment system that is composed of Drywell (DW) and Wetwell (WW) below the design pressure of 414 kPa (60 psia). In addition, the ICS continued participating in heat removal during the long-term cooling phase. A general Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach was discussed in detail relative to safety analyses of Light Water Reactor (LWR). The major components of the CSAU methodology that were highlighted particularly focused on the scaling issues of experiments and models and their applicability to the nuclear power plant transient and accidents. The major thermal-hydraulic phenomena to be analyzed were identified and the predictive models adopted in RELAP5/MOD3.3 (Patch03) code were briefly reviewed.

Nuclear reactor cavity floor passive heat removal system

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

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluidmore » communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.« less

Methods and apparatuses for deoxygenating pyrolysis oil

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

Baird, Lance Awender; Brandvold, Timothy A.; Frey, Stanley Joseph

Methods and apparatuses are provided for deoxygenating pyrolysis oil. A method includes contacting a pyrolysis oil with a deoxygenation catalyst in a first reactor at deoxygenation conditions to produce a first reactor effluent. The first reactor effluent has a first oxygen concentration and a first hydrogen concentration, based on hydrocarbons in the first reactor effluent, and the first reactor effluent includes an aromatic compound. The first reactor effluent is contacted with a dehydrogenation catalyst in a second reactor at conditions that deoxygenate the first reactor effluent while preserving the aromatic compound to produce a second reactor effluent. The second reactormore » effluent has a second oxygen concentration lower than the first oxygen concentration and a second hydrogen concentration that is equal to or lower than the first hydrogen concentration, where the second oxygen concentration and the second hydrogen concentration are based on the hydrocarbons in the second reactor effluent.« less

Methanation assembly using multiple reactors

DOEpatents

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24

A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

Status and progress of the RERTR program in the year 2003.

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

Travelli, A.; Nuclear Engineering Division

2003-01-01

One of the most important events affecting the RERTR program during the past year was the decision by the U.S. Department of Energy to request the U.S. Congress to significantly increase RERTR program funding. This decision was prompted, at least in part, by the terrible events of September 11, 2001, and by a high-level U.S./Russian Joint Expert Group recommendation to immediately accelerate RERTR program activities in both countries, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors. The U.S. Congress is expectedmore » to approve this request very soon, and the RERTR program has prepared itself well for the intense activities that the 'Accelerated RERTR Program' will require. Promising results have been obtained in the development of a fabrication process for monolithic LEU U-Mo fuel. Most existing and future research reactors could be converted to LEU with this fuel, which has a uranium density between 15.4 and 16.4 g/cm{sup 3} and yielded promising irradiation results in 2002. The most promising method hinges on producing the monolithic meat by cold-rolling a thin ingot produced by casting. The aluminum clad and the meat are bonded by friction stir welding and the cladding surface is finished by a light cold roll. This method can be applied to the production of miniplates and appears to be extendable to the production of full-size plates, possibly with intermediate anneals. Other methods planned for investigation include high temperature bonding and hot isostatic pressing. The progress achieved within the Russian RERTR program, both for the traditional tube-type elements and for the new 'universal' LEU U-Mo pin-type elements, promises to enable soon the conversion of many Russian-designed research and test reactors. Irradiation testing of both fuel types with LEU U-Mo dispersion fuels has begun. Detailed studies are in progress to define the feasibility of converting each Russian-designed research and test reactor to either fuel type. The plan for the Accelerated RERTR Program is structured to achieve LEU conversion of all HEU research reactors supplied by the United States and Russia during the next nine years. This effort will address, in addition to the fuel development and qualification, the analyses and performance/economic/safety evaluations needed to implement the conversions. In combination with this over-arching goal, the RERTR program plans to achieve at the earliest possible date qualification of LEU U-Mo dispersion fuels with uranium densities of 6 g/cm{sup 3} and 7 g/cm{sup 3}. Reactors currently using or planning to use LEU silicide fuel will rely on this fuel after termination of the FRRSNFA program, because it is acceptable to COGEMA for reprocessing. Qualification of LEU U-Mo dispersion fuels has suffered some unavoidable delays but, to accelerate it as much as possible, the RERTR program, the French CEA, and the Australian ANSTO have agreed to jointly pursue a two-element qualification test of LEU U-Mo dispersion fuel with uranium density of 7.0 g/cm{sup 3} to be performed in the Osiris reactor during 2004. The RERTR program also intends to eliminate all obstacles to the utilization of LEU in targets for isotope production, so that this important function can be performed without the need for weapons-grade materials. All of us, working together as we have for many years, can ensure that all these goals will be achieved. By promoting the efficiency and safety of research reactors while eliminating the traffic in weapons-grade uranium, we can prevent the possibility that some of this material might fall in the wrong hands. Few causes can be more deserving of our joint efforts.« less

When Do Commercial Reactors Permanently Shut Down?

EIA Publications

2011-01-01

For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

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

Code of Federal Regulations, 2014 CFR

2014-01-01

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

10 CFR 2.603 - Acceptance and docketing of application for early review of site suitability issues in a...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they are...

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

Code of Federal Regulations, 2012 CFR

2012-01-01

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

10 CFR 2.603 - Acceptance and docketing of application for early review of site suitability issues in a...

Code of Federal Regulations, 2014 CFR

2014-01-01

... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they are...

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

Code of Federal Regulations, 2013 CFR

2013-01-01

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

Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

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

Cormon, S.; Fallot, M., E-mail: fallot@subatech.in2p3.fr; Bui, V.-M.

This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {supmore » 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.« less

Bioaugmentation of activated sludge towards 3-chloroaniline removal with a mixed bacterial population carrying a degradative plasmid.

PubMed

Bathe, Stephan; Schwarzenbeck, Norbert; Hausner, Martina

2009-06-01

A bioaugmentation approach combining several strategies was applied to achieve degradation of 3-chloroaniline (3CA) in semicontinuous activated sludge reactors. In a first step, a 3CA-degrading Comamonas testosteroni strain carrying the degradative plasmid pNB2 was added to a biofilm reactor, and complete 3CA degradation together with spread of the plasmid within the indigenous biofilm population was achieved. A second set of reactors was then bioaugmented with either a suspension of biofilm cells removed from the carrier material or with biofilm-containing carrier material. 3CA degradation was established rapidly in all bioaugmented reactors, followed by a slow adaptation of the non-bioaugmented control reactors. In response to variations in 3CA concentration, all reactors exhibited temporary performance breakdowns. Whereas duplicates of the control reactors deviated in their behaviour, the bioaugmented reactors appeared more reproducible in their performance and population dynamics. Finally, the carrier-bioaugmented reactors showed an improved performance in the presence of high 3CA influent concentrations over the suspension-bioaugmented reactors. In contrast, degradation in one control reactor failed completely, but was rapidly established in the remaining control reactor.

A comparative study of thermophilic and mesophilic anaerobic co-digestion of food waste and wheat straw: Process stability and microbial community structure shifts.

PubMed

Shi, Xuchuan; Guo, Xianglin; Zuo, Jiane; Wang, Yajiao; Zhang, Mengyu

2018-05-01

Renewable energy recovery from organic solid waste via anaerobic digestion is a promising way to provide sustainable energy supply and eliminate environmental pollution. However, poor efficiency and operational problems hinder its wide application of anaerobic digestion. The effects of two key parameters, i.e. temperature and substrate characteristics on process stability and microbial community structure were studied using two lab-scale anaerobic reactors under thermophilic and mesophilic conditions. Both the reactors were fed with food waste (FW) and wheat straw (WS). The organic loading rates (OLRs) were maintained at a constant level of 3 kg VS/(m 3 ·d). Five different FW:WS substrate ratios were utilized in different operational phases. The synergetic effects of co-digestion improved the stability and performance of the reactors. When FW was mono-digested, both reactors were unstable. The mesophilic reactor eventually failed due to volatile fatty acid accumulation. The thermophilic reactor had better performance compared to mesophilic one. The biogas production rate of the thermophilic reactor was 4.9-14.8% higher than that of mesophilic reactor throughout the experiment. The shifts in microbial community structures throughout the experiment in both thermophilic and mesophilic reactors were investigated. With increasing FW proportions, bacteria belonging to the phylum Thermotogae became predominant in the thermophilic reactor, while the phylum Bacteroidetes was predominant in the mesophilic reactor. The genus Methanosarcina was the predominant methanogen in the thermophilic reactor, while the genus Methanothrix remained predominant in the mesophilic reactor. The methanogenesis pathway shifted from acetoclastic to hydrogenotrophic when the mesophilic reactor experienced perturbations. Moreover, the population of lignocellulose-degrading microorganisms in the thermophilic reactor was higher than those in mesophilic reactor, which explained the better performance of the thermophilic reactor. Copyright © 2018. Published by Elsevier Ltd.

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

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

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

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

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

Optimally moderated nuclear fission reactor and fuel source therefor

DOEpatents

Ougouag, Abderrafi M [Idaho Falls, ID; Terry, William K [Shelley, ID; Gougar, Hans D [Idaho Falls, ID

2008-07-22

An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

Flow rate analysis of wastewater inside reactor tanks on tofu wastewater treatment plant

NASA Astrophysics Data System (ADS)

Mamat; Sintawardani, N.; Astuti, J. T.; Nilawati, D.; Wulan, D. R.; Muchlis; Sriwuryandari, L.; Sembiring, T.; Jern, N. W.

2017-03-01

The research aimed to analyse the flow rate of the wastewater inside reactor tanks which were placed a number of bamboo cutting. The resistance of wastewater flow inside reactor tanks might not be occurred and produce biogas fuel optimally. Wastewater from eleven tofu factories was treated by multi-stages anaerobic process to reduce its organic pollutant and produce biogas. Biogas plant has six reactor tanks of which its capacity for waste water and gas dome was 18 m3 and 4.5 m3, respectively. Wastewater was pumped from collecting ponds to reactors by either serial or parallel way. Maximum pump capacity, head, and electrical motor power was 5m3/h, 50m, and 0.75HP, consecutively. Maximum pressure of biogas inside the reactor tanks was 55 mbar higher than atmosphere pressure. A number of 1,400 pieces of cutting bamboo at 50-60 mm diameter and 100 mm length were used as bacteria growth media inside each reactor tank, covering around 14,287 m2 bamboo area, and cross section area of inner reactor was 4,9 m2. In each reactor, a 6 inches PVC pipe was installed vertically as channel. When channels inside reactor were opened, flow rate of wastewater was 6x10-1 L.sec-1. Contrary, when channels were closed on the upper part, wastewater flow inside the first reactor affected and increased gas dome. Initially, wastewater flowed into each reactor by a gravity mode with head difference between the second and third reactor was 15x10-2m. However, head loss at the second reactor was equal to the third reactor by 8,422 x 10-4m. As result, wastewater flow at the second and third reactors were stagnant. To overcome the problem pump in each reactor should be installed in serial mode. In order to reach the output from the first reactor and the others would be equal, and biogas space was not filled by wastewater, therefore biogas production will be optimum.

Operators in the Plum Brook Reactor Facility Control Room

NASA Image and Video Library

1970-03-21

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

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Reactors, the Director, Office of Nuclear Reactor Regulation, the Director, Office of Nuclear Material... Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and State... be requested to: (i) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office...

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Reactors, the Director, Office of Nuclear Reactor Regulation, the Director, Office of Nuclear Material... Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and State... be requested to: (i) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office...

Weld monitor and failure detector for nuclear reactor system

DOEpatents

Sutton, Jr., Harry G.

1987-01-01

Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

Comparison of Calibration of Sensors Used for the Quantification of Nuclear Energy Rate Deposition

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

Brun, J.; Reynard-Carette, C.; Tarchalski, M.

This present work deals with a collaborative program called GAMMA-MAJOR 'Development and qualification of a deterministic scheme for the evaluation of GAMMA heating in MTR reactors with exploitation as example MARIA reactor and Jules Horowitz Reactor' between the National Centre for Nuclear Research of Poland, the French Atomic Energy and Alternative Energies Commission and Aix Marseille University. One of main objectives of this program is to optimize the nuclear heating quantification thanks to calculation validated from experimental measurements of radiation energy deposition carried out in irradiation reactors. The quantification of the nuclear heating is a key data especially for themore » thermal, mechanical design and sizing of irradiation experimental devices in specific irradiated conditions and locations. The determination of this data is usually performed by differential calorimeters and gamma thermometers such as used in the experimental multi-sensors device called CARMEN 'Calorimetric en Reacteur et Mesures des Emissions Nucleaires'. In the framework of the GAMMA-MAJOR program a new calorimeter was designed for the nuclear energy deposition quantification. It corresponds to a single-cell calorimeter and it is called KAROLINA. This calorimeter was recently tested during an irradiation campaign inside MARIA reactor in Poland. This new single-cell calorimeter differs from previous CALMOS or CARMEN type differential calorimeters according to three main points: its geometry, its preliminary out-of-pile calibration, and its in-pile measurement method. The differential calorimeter, which is made of two identical cells containing heaters, has a calibration method based on the use of steady thermal states reached by simulating the nuclear energy deposition into the calorimeter sample by Joule effect; whereas the single-cell calorimeter, which has no heater, is calibrated by using the transient thermal response of the sensor (heating and cooling steps). The paper will concern these two kinds of calorimetric sensors. It will focus in particular on studies on their out-of-pile calibrations. Firstly, the characteristics of the sensor designs will be detailed (such as geometry, dimension, material sample, assembly, instrumentation). Then the out-of-pile calibration methods will be described. Furthermore numerical results obtained thanks to 2D axisymmetrical thermal simulations (Finite Element Method, CAST3M) and experimental results will be presented for each sensor. A comparison of the two different thermal sensor behaviours will be realized. To conclude a discussion of the advantages and the drawbacks of each sensor will be performed especially regarding measurement methods. (authors)« less

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

NASA Technical Reports Server (NTRS)

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

2016-01-01

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

Reactor water cleanup system

DOEpatents

Gluntz, Douglas M.; Taft, William E.

1994-01-01

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

The role of nuclear reactors in space exploration and development

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

Lipinski, R.J.

2000-07-01

The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 235}U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. Onemore » reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new frontier for their children and their grandchildren. They pave the way for all life on earth to move out into the solar system. At one time, humans built and flew space reactors; it is time to do so again.« less

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

NASA Astrophysics Data System (ADS)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were elucidated in light of the analyzed degradation products.

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2011 CFR

2011-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

Unmixed fuel processors and methods for using the same

DOEpatents

Kulkarni, Parag Prakash; Cui, Zhe

2010-08-24

Disclosed herein are unmixed fuel processors and methods for using the same. In one embodiment, an unmixed fuel processor comprises: an oxidation reactor comprising an oxidation portion and a gasifier, a CO.sub.2 acceptor reactor, and a regeneration reactor. The oxidation portion comprises an air inlet, effluent outlet, and an oxygen transfer material. The gasifier comprises a solid hydrocarbon fuel inlet, a solids outlet, and a syngas outlet. The CO.sub.2 acceptor reactor comprises a water inlet, a hydrogen outlet, and a CO.sub.2 sorbent, and is configured to receive syngas from the gasifier. The regeneration reactor comprises a water inlet and a CO.sub.2 stream outlet. The regeneration reactor is configured to receive spent CO.sub.2 adsorption material from the gasification reactor and to return regenerated CO.sub.2 adsorption material to the gasification reactor, and configured to receive oxidized oxygen transfer material from the oxidation reactor and to return reduced oxygen transfer material to the oxidation reactor.

Thermionic switched self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

1989-01-01

A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, R.B.; Fero, A.H.; Sejvar, J.

1997-12-16

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

1997-01-01

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

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

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

Propellant actuated nuclear reactor steam depressurization valve

DOEpatents

Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

1992-01-01

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

DOEpatents

West, J.M.; Weills, J.T.

1960-03-15

A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

Thermal-hydraulic interfacing code modules for CANDU reactors

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

Liu, W.S.; Gold, M.; Sills, H.

1997-07-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

97. ARAIII. ML1 reactor has been moved into GCRE reactor ...

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

97. ARA-III. ML-1 reactor has been moved into GCRE reactor building (ARA-608) for examination of corrosion on its underside and repair. May 24, 1963. Ineel photo no. 63-3485. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

NEUTRONIC REACTOR MANIPULATING DEVICE

DOEpatents

Ohlinger, L.A.

1962-08-01

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

Trade studies for nuclear space power systems

NASA Technical Reports Server (NTRS)

Smith, John M.; Bents, David J.; Bloomfield, Harvey S.

1991-01-01

As human visions of space applications expand and as we probe further out into the universe, our needs for power will also expand, and missions will evolve which are enabled by nuclear power. A broad spectrum of missions which are enhanced or enabled by nuclear power sources have been defined. These include Earth orbital platforms, deep space platforms, planetary exploration, and terrestrial resource exploration. The recently proposed Space Exploration Initiative (SEI) to the Moon and Mars has more clearly defined these missions and their power requirements. Presented here are results of recent studies of radioisotope and nuclear reactor energy sources, combined with various energy conversion devices for Earth orbital applications, SEI lunar/Mars rovers, surface power, and planetary exploration.

Pen Branch stream corridor and Delta Wetlands change assessment

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

Blohm, J.D.

1995-06-01

Airborne multispectral scanner data from 1987 to 1991 covering the Pen Branch corridor and delta at SRS were utilized to provide a detailed change detection analysis. The multispectral data were geo-referenced to a Universal Transverse Mercator projection using finite element registration. Each year was then classified into eleven different landcover categories, and the yearly changes in each landcover category were analyzed. The decrease in operations of K Reactor in 1988 has resulted in drying of the corridor and delta. This has led to the decline of nonpersistent vegetation and the increase of persistent vegetation. Cattails, willow, and bottomland hardwoods, inmore » particular, have grown to dominate the corridor and most of the delta.« less

Instrumental neutron activation analysis data for cloud-water particulate samples, Mount Bamboo, Taiwan

USGS Publications Warehouse

Lin, Neng-Huei; Sheu, Guey-Rong; Wetherbee, Gregory A.; Debey, Timothy M.

2013-01-01

Cloud water was sampled on Mount Bamboo in northern Taiwan during March 22-24, 2002. Cloud-water samples were filtered using 0.45-micron filters to remove particulate material from the water samples. Filtered particulates were analyzed by instrumental neutron activation analysis (INAA) at the U.S. Geological Survey National Reactor Facility in Denver, Colorado, in February 2012. INAA elemental composition data for the particulate materials are presented. These data complement analyses of the aqueous portion of the cloud-water samples, which were performed earlier by the Department of Atmospheric Sciences, National Central University, Taiwan. The data are intended for evaluation of atmospheric transport processes and air-pollution sources in Southeast Asia.

40 CFR 63.1406 - Reactor batch process vent provisions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 11 2011-07-01 2011-07-01 false Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-09

... Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft NUREG; request for comment. SUMMARY: The U.S..., Revision 10, ``Operator Licensing Examination Standards for Power Reactors.'' DATES: Submit comments [email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...

76 FR 55718 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels The ACRS Subcommittee on Materials, Metallurgy & Reactor...'' for reactor coolant system (RCS) components, as mentioned in 10 CFR 50 Appendix A, GDC-4. The...

40 CFR 63.1406 - Reactor batch process vent provisions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

75 FR 58449 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-24

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels The ACRS Subcommittee on Materials, Metallurgy & Reactor... would result in a major inconvenience. Dated: September 17, 2010. Antonio Dias, Chief, Reactor Safety...

151. ARAIII Reactor building (ARA608) Details of reactor pit and ...

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

151. ARA-III Reactor building (ARA-608) Details of reactor pit and instrument plan. Aerojet-general 880-area/GCRE-608-T-19. Date: November 1958. Ineel index code no. 063-0608-25-013-102678. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor... be designed to store spent fuel and/or solid reactor-related GTCC waste. (1) Reactor-related GTCC...

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

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

A Review of Gas-Cooled Reactor Concepts for SDI Applications

DTIC Science & Technology

1989-08-01

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

Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

NASA Technical Reports Server (NTRS)

Tew, R. C.; Jefferies, K. S.

1974-01-01

A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

Numerical Simulations of a 96-rod Polysilicon CVD Reactor

NASA Astrophysics Data System (ADS)

Guoqiang, Tang; Cong, Chen; Yifang, Cai; Bing, Zong; Yanguo, Cai; Tihu, Wang

2018-05-01

With the rapid development of the photovoltaic industry, pressurized Siemens belljar-type polysilicon CVD reactors have been enlarged from 24 rods to 96 rods in less than 10 years aimed at much greater single-reactor productivity. A CFD model of an industry-scale 96-rod CVD reactor was established to study the internal temperature distribution and the flow field of the reactor. Numerical simulations were carried out and compared with actual growth results from a real CVD reactor. Factors affecting polysilicon depositions such as inlet gas injections, flow field, and temperature distribution in the CVD reactor are studied.

Reactor water cleanup system

DOEpatents

Gluntz, D.M.; Taft, W.E.

1994-12-20

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

Characteristics and Dose Levels for Spent Reactor Fuels

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

Coates, Cameron W

2007-01-01

Current guidance considers highly radioactive special nuclear materials to be those materials that, unshielded, emit a radiation dose [rate] measured at 1 m which exceeds 100 rem/h. Smaller, less massive fuel assemblies from research reactors can present a challenge from the point of view of self protection because of their size (lower dose, easier to handle) and the desirability of higher enrichments; however, a follow-on study to cross-compare dose trends of research reactors and power reactors was deemed useful to confirm/verify these trends. This paper summarizes the characteristics and dose levels of spent reactor fuels for both research reactors andmore » power reactors and extends previous studies aimed at quantifying expected dose rates from research reactor fuels worldwide.« less

Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

NASA Astrophysics Data System (ADS)

Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

2017-02-01

In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

Reactor operation environmental information document

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

Haselow, J.S.; Price, V.; Stephenson, D.E.

1989-12-01

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimalmore » impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.« less

Auxiliary reactor for a hydrocarbon reforming system

DOEpatents

Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

2006-01-17

An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

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

Lister, Tedd E; Parkman, Jacob A; Diaz Aldana, Luis A

A method of recovering metals from electronic waste comprises providing a powder comprising electronic waste in at least a first reactor and a second reactor and providing an electrolyte comprising at least ferric ions in an electrochemical cell in fluid communication with the first reactor and the second reactor. The method further includes contacting the powders within the first reactor and the second reactor with the electrolyte to dissolve at least one base metal from each reactor into the electrolyte and reduce at least some of the ferric ions to ferrous ions. The ferrous ions are oxidized at an anodemore » of the electrochemical cell to regenerate the ferric ions. The powder within the second reactor comprises a higher weight percent of the at least one base metal than the powder in the first reactor. Additional methods of recovering metals from electronic waste are also described, as well as an apparatus of recovering metals from electronic waste.« less

High-intensity power-resolved radiation imaging of an operational nuclear reactor.

PubMed

Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

2015-10-09

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

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.

High-intensity power-resolved radiation imaging of an operational nuclear reactor

PubMed Central

Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

2015-01-01

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho; Lee, Sangjin; Jin, Yoon-Su; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2014-09-01

This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

A comparison of the technological effectiveness of dairy wastewater treatment in anaerobic UASB reactor and anaerobic reactor with an innovative design.

PubMed

Jedrzejewska-Cicinska, M; Kozak, K; Krzemieniewski, M

2007-10-01

The present research was an investigation of the influence of an innovative design of reactor filled with polyethylene (PE) granulate on model dairy wastewater treatment efficiency under anaerobic conditions compared to that obtained in a typical UASB reactor. The experiment was conducted at laboratory scale. An innovative reactor was designed with the reaction chamber inclined 30 degrees in relation to the ground with upward waste flow and was filled with PE granular material. Raw model dairy wastewater was fed to two anaerobic reactors of different design at the organic loading rate of 4 kg COD m(-3)d(-1). Throughout the experiment, a higher removal efficiency of organic compounds was observed in the reactor with an innovative design and it was higher by 7.1% on average than in the UASB reactor. The total suspended solids was lower in the wastewater treated in the anaerobic reactor with the innovative design. Applying a PE granulated filling in the chamber of the innovative reactor contributed to an even distribution of sludge biomass in the reactor, reducing washout of anaerobic sludge biomass from the reaction chamber and giving a higher organic compounds removal efficiency.

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Development of toroid-type HTS DC reactor series for HVDC system

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-11-01

This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Experiences in utilization of research reactors in Yugoslavia

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

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

1971-06-15

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

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.

SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

DOEpatents

Ransohoff, J.A.; Plawchan, J.D.

1960-08-16

A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

DOEpatents

Loeb, E.

1961-01-17

A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

System and method for air temperature control in an oxygen transport membrane based reactor

DOEpatents

Kelly, Sean M

2016-09-27

A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

System and method for temperature control in an oxygen transport membrane based reactor

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

Kelly, Sean M.

A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2010-09-29

to design a smaller scale version of a naval pressurized water reactor , or to design a new reactor type potentially using a thorium liquid salt...integrated nuclear power system capable of use on destroyer- sized vessels either using a pressurized water reactor or a thorium liquid salt reactor ...nuclear reactors for Navy surface ships. The text of Section 246 is as follows: SEC. 246. STUDY ON THORIUM -LIQUID FUELED REACTORS FOR NAVAL FORCES

Modification of UASB reactor by using CFD simulations for enhanced treatment of municipal sewage.

PubMed

Das, Suprotim; Sarkar, Supriya; Chaudhari, Sanjeev

2018-02-01

Up-flow anaerobic sludge blanket (UASB) has been in use since last few decades for the treatment of organic wastewaters. However, the performance of UASB reactor is quite low for treatment of low strength wastewaters (LSWs) due to less biogas production leading to poor mixing. In the present research work, a modification was done in the design of UASB to improve mixing of reactor liquid which is important to enhance the reactor performance. The modified UASB (MUASB) reactor was designed by providing a slanted baffle along the height of the reactor having an angle of 5.7° with the vertical wall. A two-dimensional computational fluid dynamics (CFD) simulation of three phase gas-liquid-solid flow in MUASB reactor was performed and compared with conventional UASB reactor. The CFD study indicated better mixing in terms of vorticity magnitude in MUASB reactor as compared to conventional UASB, which was reflected in the reactor performance. The performance of MUASB was compared with conventional UASB reactor for the onsite treatment of domestic sewage as LSW. Around 16% higher total chemical oxygen demand removal efficiency was observed in MUASB reactor as compared to conventional UASB during this study. Therefore, this MUASB model demonstrates a qualitative relationship between mixing and performance during the treatment of LSW. From the study, it seems that MUASB holds promise for field applications.

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2010 CFR

2010-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2011 CFR

2011-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

78 FR 26811 - Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-08

... Research Reactor; License Renewal for the Dow Chemical TRIGA Research Reactor; Supplemental Information and... 20, 2012 (77 FR 42771), ``License Renewal for the Dow Chemical TRIGA Research Reactor,'' to inform... Chemical Company which would authorize continued operation of the Dow TRIGA Research Reactor. The notice...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Amount of financial protection required for other reactors... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Amount of financial protection required for other reactors... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000...

PBF Reactor Building (PER620). Camera faces north into highbay/reactor pit ...

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

PBF Reactor Building (PER-620). Camera faces north into high-bay/reactor pit area. Inside from for reactor enclosure is in place. Photographer: John Capek. Date: March 15, 1967. INEEL negative no. 67-1769 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Reactor safety method

DOEpatents

Vachon, Lawrence J.

1980-03-11

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

155. ARAIII Reactor building (ARA608) Details of reactor pit showing ...

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

155. ARA-III Reactor building (ARA-608) Details of reactor pit showing tray supports and fuel element storage rack. Aerojet-general 880-area/GCRE-608-MS-2. Date: November 1958. Ineel index code no. 063-0608-40-013-102625. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

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

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

Not Available

1979-06-01

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

Breeder Reactors, Understanding the Atom Series.

ERIC Educational Resources Information Center

Mitchell, Walter, III; Turner, Stanley E.

The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 2.103 - Action on applications for byproduct, source, special nuclear material, facility and operator...

Code of Federal Regulations, 2011 CFR

2011-01-01

..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear... of this chapter, see § 2.106(d). (b) If the Director, Office of Nuclear Reactor Regulation, Director...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 2.103 - Action on applications for byproduct, source, special nuclear material, facility and operator...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear... of this chapter, see § 2.106(d). (b) If the Director, Office of Nuclear Reactor Regulation, Director...

Computer model of catalytic combustion/Stirling engine heater head

NASA Technical Reports Server (NTRS)

Chu, E. K.; Chang, R. L.; Tong, H.

1981-01-01

The basic Acurex HET code was modified to analyze specific problems for Stirling engine heater head applications. Specifically, the code can model: an adiabatic catalytic monolith reactor, an externally cooled catalytic cylindrical reactor/flat plate reactor, a coannular tube radiatively cooled reactor, and a monolithic reactor radiating to upstream and downstream heat exchangers.

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

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

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

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

Cao, Yan; Gohar, Yousry

2015-11-01

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

Pressurized fluidized bed reactor

DOEpatents

Isaksson, J.

1996-03-19

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Pressurized fluidized bed reactor

DOEpatents

Isaksson, Juhani

1996-01-01

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Microdosimetric investigation of the spectra from YAYOI by use of the Monte Carlo code PHITS.

PubMed

Nakao, Minoru; Baba, Hiromi; Oishi, Ayumu; Onizuka, Yoshihiko

2010-07-01

The purpose of this study was to obtain the neutron energy spectrum on the surface of the moderator of the Tokyo University reactor YAYOI and to investigate the origins of peaks observed in the neutron energy spectrum by use of the Monte Carlo Code PHITS for evaluating biological studies. The moderator system was modeled with the use of details from an article that reported a calculation result and a measurement result for a neutron spectrum on the surface of the moderator of the reactor. Our calculation results with PHITS were compared to those obtained with the discrete ordinate code ANISN described in the article. In addition, the changes in the neutron spectrum at the boundaries of materials in the moderator system were examined with PHITS. Also, microdosimetric energy distributions of secondary charged particles from neutron recoil or reaction were calculated by use of PHITS and compared with a microdosimetric experiment. Our calculations of the neutron energy spectrum with PHITS showed good agreement with the results of ANISN in terms of the energy and structure of the peaks. However, the microdosimetric dose distribution spectrum with PHITS showed a remarkable discrepancy with the experimental one. The experimental spectrum could not be explained by PHITS when we used neutron beams of two mono-energies.

Qualitative Analysis of Microbial Dynamics during Anaerobic Digestion of Microalgal Biomass in a UASB Reactor

PubMed Central

Doloman, Anna; Soboh, Yousef; Walters, Andrew J.; Sims, Ronald C.

2017-01-01

Anaerobic digestion (AD) is a microbiologically coordinated process with dynamic relationships between bacterial players. Current understanding of dynamic changes in the bacterial composition during the AD process is incomplete. The objective of this research was to assess changes in bacterial community composition that coordinates with anaerobic codigestion of microalgal biomass cultivated on municipal wastewater. An upflow anaerobic sludge blanket reactor was used to achieve high rates of microalgae decomposition and biogas production. Samples of the sludge were collected throughout AD and extracted DNA was subjected to next-generation sequencing using methanogen mcrA gene specific and universal bacterial primers. Analysis of the data revealed that samples taken at different stages of AD had varying bacterial composition. A group consisting of Bacteroidales, Pseudomonadales, and Enterobacteriales was identified to be putatively responsible for the hydrolysis of microalgal biomass. The methanogenesis phase was dominated by Methanosarcina mazei. Results of observed changes in the composition of microbial communities during AD can be used as a road map to stimulate key bacterial species identified at each phase of AD to increase yield of biogas and rate of substrate decomposition. This research demonstrates a successful exploitation of methane production from microalgae without any biomass pretreatment. PMID:29259629

Manufacturing Technology of Ceramic Pebbles for Breeding Blanket

PubMed Central

Stefanelli, Eleonora; Del Serra, Daniele; Malquori, Stefano

2018-01-01

An open issue for the fusion power reactor is the choice of breeding blanket material. The possible use of Helium-Cooled Pebble Breeder ceramic material in the form of pebble beds is of great interest worldwide as demonstrated by the numerous studies and research on this subject. Lithium orthosilicate (Li4SiO4) is a promising breeding material investigated in this present study because the neutron capture of Li-6 allows the production of tritium, 6Li (n, t) 4He. Furthermore, lithium orthosilicate has the advantages of low activation characteristics, low thermal expansion coefficient, high thermal conductivity, high density and stability. Even if they are far from the industrial standard, a variety of industrial processes have been proposed for making orthosilicate pebbles with diameters of 0.1–1 mm. However, some manufacturing problems have been observed, such as in the chemical stability (agglomeration phenomena). The aim of this study is to provide a new methodology for the production of pebbles based on the drip casting method, which was jointly developed by the DICI-University of Pisa and Industrie Bitossi. Using this new (and alternative) manufacturing technology, in the field of fusion reactors, appropriately sized ceramic pebbles could be produced for use as tritium breeders. PMID:29724071

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

Hannan, N. A.; Matos, J. E.; Stillman, J. A.

At the request of the Czech Technical University (CTU) in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. Fore core configurations C1 and C2, criticality calculations were done for cases with all controlmore » rods at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were doe for the C1 core configuration. The reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations. Finally, the reactivity feedback coefficients, the prompt neutron lifetime, and the total effective delay neutron fraction were calculated for each of the three cores.« less

Scaling the Shear-flow Stabilized Z-pinch to Reactor Conditions

NASA Astrophysics Data System (ADS)

McLean, H. S.; Schmidt, A.; Shumlak, U.; Nelson, B. A.; Golingo, R. P.; Cleveau, E.

2015-11-01

We present a conceptual design along with scaling calculations for a pulsed fusion reactor based on the shear-flow-stabilized Z-pinch device. Experiments performed on the ZaP device, at the University of Washington, have demonstrated stable operation for durations of 20 usec at ~100kA discharge current for pinches that are ~1 cm in diameter and 100 cm long. The inverse of the pinch diameter and plasma energy density scale strongly with pinch current and calculations show that maintaining stabilization durations of ~7 usec for increased discharge current (~15x) in a shortened pinch (10 cm) results in a pinch diameter of ~200 um and plasma conditions that approach those needed to support significant fusion burn and energy gain (Ti ~ 30keV, density ~ 3e26/m3, ntau ~1.4e20 sec/m3). Compelling features of the concept include operation at modest discharge current (1.5 MA) and voltage (40kV) along with direct adoption of liquid metals for at least one electrode--technological capabilities that have been proven in existing, commercial, pulse power devices such as large ignitrons. LLNL-ABS-674920. This work performed under the auspices of the U.S. Department of Energy ARPAe ALPHA Program by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Weak-beam scanning transmission electron microscopy for quantitative dislocation density measurement in steels.

PubMed

Yoshida, Kenta; Shimodaira, Masaki; Toyama, Takeshi; Shimizu, Yasuo; Inoue, Koji; Yoshiie, Toshimasa; Milan, Konstantinovic J; Gerard, Robert; Nagai, Yasuyoshi

2017-04-01

To evaluate dislocations induced by neutron irradiation, we developed a weak-beam scanning transmission electron microscopy (WB-STEM) system by installing a novel beam selector, an annular detector, a high-speed CCD camera and an imaging filter in the camera chamber of a spherical aberration-corrected transmission electron microscope. The capabilities of the WB-STEM with respect to wide-view imaging, real-time diffraction monitoring and multi-contrast imaging are demonstrated using typical reactor pressure vessel steel that had been used in an European nuclear reactor for 30 years as a surveillance test piece with a fluence of 1.09 × 1020 neutrons cm-2. The quantitatively measured size distribution (average loop size = 3.6 ± 2.1 nm), number density of the dislocation loops (3.6 × 1022 m-3) and dislocation density (7.8 × 1013 m m-3) were carefully compared with the values obtained via conventional weak-beam transmission electron microscopy studies. In addition, cluster analysis using atom probe tomography (APT) further demonstrated the potential of the WB-STEM for correlative electron tomography/APT experiments. © The Author 2017. Published by Oxford University Press on behalf of The Japanese Society of Microscopy. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

Reactor vibration reduction based on giant magnetostrictive materials

NASA Astrophysics Data System (ADS)

Rongge, Yan; Weiying, Liu; Yuechao, Wu; Menghua, Duan; Xiaohong, Zhang; Lihua, Zhu; Ling, Weng; Ying, Sun

2017-05-01

The vibration of reactors not only produces noise pollution, but also affects the safe operation of reactors. Giant magnetostrictive materials can generate huge expansion and shrinkage deformation in a magnetic field. With the principle of mutual offset between the giant magnetostrictive force produced by the giant magnetostrictive material and the original vibration force of the reactor, the vibration of the reactor can be reduced. In this paper, magnetization and magnetostriction characteristics in silicon steel and the giant magnetostrictive material are measured, respectively. According to the presented magneto-mechanical coupling model including the electromagnetic force and the magnetostrictive force, reactor vibration is calculated. By comparing the vibration of the reactor with different inserted materials in the air gaps between the reactor cores, the vibration reduction effectiveness of the giant magnetostrictive material is validated.

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

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

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

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

Solution of heat removal from nuclear reactors by natural convection

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav

2014-03-01

This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR).The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor) for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

Imaging Fukushima Daiichi reactors with muons

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

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

2013-05-15

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

Imaging Fukushima Daiichi reactors with muons

NASA Astrophysics Data System (ADS)

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

2013-05-01

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

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

Fitzpatrick, F.C.; Gray, D.D.; Hyndman, J.R.

The thermal, ecological, and social impacts of a 40-reactor NEC are compared to impacts from four 10-reactor NECs and ten 4-reactor power plants. The comparison was made for surrogate sites in western Tennessee. The surrogate site for the 40-reactor NEC is located on Kentucky Lake. A layout is postulated for ten clusters of four reactors each with 2.5-mile spacing between clusters. The plants use natural-draft cooling towers. A transmission system is proposed for delivering the power (48,000 MW) to five load centers. Comparable transmission systems are proposed for the 10-reactor NECs and the 4-reactor dispersed sites delivering power to themore » same load centers. (auth)« less