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Sample records for advanced test reactor critical facility

  1. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    International Nuclear Information System (INIS)

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830

  2. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  3. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  4. The Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Todd R. Allen; Collin J. Knight; Jeff B. Benson; Frances M. Marshall; Mitchell K. Meyer; Mary Catherine Thelen

    2011-08-01

    In 2007, the Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), was designated by the Department of Energy (DOE) as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by approved researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide those researchers with the best ideas access to the most advanced test capability, regardless of the proposer’s physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, obtained access to additional PIE equipment, taken steps to enable the most advanced post-irradiation analysis possible, and initiated an educational program and digital learning library to help potential users better understand the critical issues in reactor technology and how a test reactor facility could be used to address this critical research. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program invited universities to nominate their capability to become part of a broader user facility. Any university is eligible to self-nominate. Any nomination is then peer reviewed to ensure that the addition of the university facilities adds useful capability to the NSUF. Once added to the NSUF team, the university capability is then integral to the NSUF operations and is available to all users via the proposal process. So far, six universities have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these university capabilities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user’s technical needs. The current NSUF partners are

  5. Advanced Test Reactor National Scientific User Facility Partnerships

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

    2012-03-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of

  6. Advanced Test Reactor National Scientific User Facility Partnerships

    International Nuclear Information System (INIS)

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

  7. The Advanced Test Reactor as a National Scientific User Facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) has been in operation since 1967 and mainly used to support U.S. Department of Energy (US DOE) materials and fuels research programs. Irradiation capabilities of the ATR and post-irradiation examination capabilities of the Idaho National Laboratory (INL) were generally not being utilized by universities and other potential users due largely to a prohibitive pricing structure. While materials and fuels testing programs using the ATR continue to be needed for US DOE programs such as the Advanced Fuel Cycle Initiative and Next Generation Nuclear Plant, US DOE recognized there was a national need to make these capabilities available to a broader user base. In April 2007, the U.S. Department of Energy designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). As a NSUF, most of the services associated with university experiment irradiation and post-irradiation examinations are provided free-of-charge. The US DOE is providing these services to support U.S. leadership in nuclear science, technology, and education and to encourage active university/industry/laboratory collaboration. The first full year of implementing the user facility concept was 2008 and it was a very successful year. The first university experiment pilot project was developed in collaboration with the University of Wisconsin and began irradiation in the ATR in 2008. Lessons learned from this pilot program will be applied to future NSUF projects. Five other university experiments were also competitively selected in March 2008 from the initial solicitation for proposals. The NSUF now has a continually open process where universities can submit proposals as they are ready. Plans are to invest in new and upgraded capabilities at the ATR, post-irradiation examination capabilities at the INL, and in a new experiment assembly facility to further support the implementation of the user facility concept. Through a newly created Partnership Program

  8. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  9. Advanced Test Reactor National Scientific User Facility Progress

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to 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, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  10. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    International Nuclear Information System (INIS)

    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 is 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

  11. The advanced test reactor national scientific user facility advancing nuclear technology

    International Nuclear Information System (INIS)

    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 is 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

  12. The advanced test reactor national scientific user facility: advancing nuclear technology education

    International Nuclear Information System (INIS)

    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 designated the Idaho National Laboratory (INL) Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The ATR NSUF provides education programs including a Users Week, internships, faculty student team projects and faculty/staff exchanges. In addition, the 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. (author)

  13. Advancing nuclear technology and research. The advanced test reactor national scientific user facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research. The mission of the ATR NSUF is to provide access to world-class facilities, thereby facilitating the advancement of nuclear science and technology. Cost free access to the ATR, INL post irradiation examination facilities, and partner facilities is 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 United States Department of Energy. 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. (author)

  14. Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

    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 groups 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  16. Critical need for MFE: the Alcator DX advanced divertor test facility

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Wolf, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

    2013-10-01

    Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: how to (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as ``Super X'' and ``X-point target'' may allow a fully detached, low temperature plasma to be produced in the divertor while maintaining a hot boundary layer around a clean plasma core - a potential game-changer for magnetic fusion. No facility currently exists to test these ideas at the required parallel heat flux densities. Alcator DX will be a national facility, employing the high magnetic field technology of Alcator combined with high-power ICRH and LHCD to test advanced divertor concepts at FNSF/DEMO power exhaust densities and plasma pressures. Its extended vacuum vessel contains divertor cassettes with poloidal field coils for conventional, snowflake, super-X and X-point target geometries. Divertor and core plasma performance will be explored in regimes inaccessible in conventional devices. Reactor relevant ICRF and LH drivers will be developed, utilizing high-field side launch platforms for low PMI. Alcator DX will inform the conceptual development and accelerate the readiness-for-deployment of next-step fusion facilities.

  17. Research on reactor physics using the Japan Materials Testing Reactor Critical Facility (JMTRC)

    International Nuclear Information System (INIS)

    The JMTRC of 100 W was installed for the purpose of carrying out the basic experiment on the nuclear characteristics of reactors and the preceding test related to the operation plan of the Japan material testing reactor (JMTR, 50 MW). After the attainment of the initial criticality in October, 1965, for obtaining the reactor physics characteristics, criticality experiment was begun. The items of the criticality experiment were critical mass, control rod worth, reactor dynamic characteristic parameters, shutdown margin and so on, and these experimental data were effectively utilized for the safety evaluation in the operation of the JMTR. The preceding test using the JMTRC has been carried out for obtaining the nuclear characteristics of samples and the thermal characteristics estimated from those results by simulating the JMTR core. In August, 1983, the degree of fuel enrichment for the JMTRC was reduced to 45 % U-235, and various experiments usig the MEU core were carried out. In this paper, the criticality experiment using the MEU core and the experiment on the characteristics of lithium-containing pellets are reported. (K.I.)

  18. Advanced Test Reactor National Scientific User Facility 2010 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Mary Catherine Thelen; Todd R. Allen

    2011-05-01

    This is the 2010 ATR National Scientific User Facility Annual Report. This report provides an overview of the program for 2010, along with individual project reports from each of the university principal investigators. The report also describes the capabilities offered to university researchers here at INL and at the ATR NSUF partner facilities.

  19. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  20. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition as part of a fuel meat thickness optimization effort for reactor performance other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

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

    Energy Technology Data Exchange (ETDEWEB)

    Dan Ogden

    2014-10-01

    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 the 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Soelberg, Renae [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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 the 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

  3. Advanced Test Reactor Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego; Brion Bennett

    2011-11-01

    U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Advanced Test Reactor Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. U.S. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

  4. Advanced test reactor. Testing capabilities and plans

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plants for the NSUF. (author)

  5. Advanced Control Test Operation (ACTO) facility

    International Nuclear Information System (INIS)

    The Advanced Control Test Operation (ACTO) project, sponsored by the US Department of Energy (DOE), is being developed to enable the latest modern technology, automation, and advanced control methods to be incorporated into nuclear power plants. The facility is proposed as a national multi-user center for advanced control development and testing to be completed in 1991. The facility will support a wide variety of reactor concepts, and will be used by researchers from Oak Ridge National Laboratory (ORNL), plus scientists and engineers from industry, other national laboratories, universities, and utilities. ACTO will also include telecommunication facilities for remote users

  6. Testing of the Y-12 Plant Criticality Accident Alarm System detectors at the Sandia Pulsed Reactor Facility

    International Nuclear Information System (INIS)

    The Oak Ridge Y-12 Plant operates its Criticality Accident Alarm System (CAAS) according to the guidance of Standard ANSI/ANS-8.3-1986. This standard requires that the detector shall not fail to initiate an alarm when subjected to a radiation field of at least 0.1 Gy/s (10 rad/s). It also requires that the system shall be designed to immediately detect the minimum accident of concern and shall produce an alarm within one half second of activation. Sixty-three new detectors that use plastic scintillators have been obtained to upgrade the current Y-12 Plant CAAS. To ensure that these detectors can support the above criteria, testing was done using the SPR III reactor at the Sandia Pulsed Reactor Facility

  7. New Sensors for In-Pile Temperature Detection at the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; J. E. Daw; K. G. Condie; S. Curtis Wilkins

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation’s energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

  8. New Sensors for In-Pile Temperature Detection at the Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation's energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

  9. The key role of critical mock-up facilities for neutronic physics assessment of advanced reactors: an overview of Cea Cadarache tools

    International Nuclear Information System (INIS)

    The Experimental Physics section of CEA Cadarache operates three critical facilities devoted to neutronic studies of advanced reactors (EOLE, MINERVE and MASURCA) covering a large scope of interests. These include 100% MOX core in ABWR qualification, knowledge improvement of basic nuclear data for heavy nuclides for new options of the fuel cycle - especially the multi-recycling of plutonium - and accelerator-driven systems neutronic behaviour for transmutation studies. The paper describes these facilities, the scientific programmes associated and the progressive improvement of experimental techniques, the aim being to significantly reduce the uncertainties regarding the evaluation of the physical parameters. (authors)

  10. Westinghouse critical heat flux test facility. ODEN

    Energy Technology Data Exchange (ETDEWEB)

    Majed, Mahdi; Andersson, Stig; Waldemarsson, Fredrik [Westinghouse Electric Sweden AB, Vaesteraas (Sweden)

    2012-11-01

    Westinghouse has designed and built ODEN, a new Critical Heat Flux test loop for PWR applications. The ODEN test facility is a replacement to (and improvement upon) the well known former Heat Transfer Research Facility (HTRF) of Columbia University in New York City. The ODEN loop shares the lab infrastructure (power supply, heat sink and control room) with the well-known FRIGG BRW test loop. The ODEN loop is designed to cover DNB testing needs for all types of PWR lattices in 5x5 or 6x6 rectangular geometry or in hexagonal test sections. The loop installation was completed in 2006, shakedown testing in 2009, and qualification / benchmark testing versus HTRF data was completed in 2010. The qualification test and the benchmark results are reported in [1] and [2]. The ODEN critical heat flux test loop has been utilized recently to perform DNB measurements on the Westinghouse fuel design for VVER 1000 type reactors. The test bundle configuration is a 19 rod hexagonal array. The fuel has been tested in an extensive thermal-hydraulic verification program with axially uniform test (typical cell) and two axially cosine tests (typical and thimble cells). The DNB measurements have been performed at low to high pressures, low to high mass flows and includes very high steam quality conditions. The ODEN loop has again showed the high DNB data quality, and excellent agreement of the repeatability of the DNB data. (orig.)

  11. Test Facility for SMART Reactor Flow Distribution

    International Nuclear Information System (INIS)

    A Reactor Flow Distribution Test Facilities for SMART, named SCOP (SMART Core Flow and Pressure Test Facility), were designed in order to simulate the distributions of (1) core flow and (2) reactor sectional flow resistance and flow rates. SCOP facility was designed based on the linear scaling law in order to preserve the flow characteristics of the prototype system, which are distributions of flow rate and pressure drop. The reduced scale was selected as a 1/5 of prototype length scale. The nominal flow condition was designed to be similar based on the velocity as that of the SMART reactor, which can minimize the flow distortion in the reduced scale of test facility by maintaining high Re number flow. Test facility includes fluid system, control/instrumentation system, data acquisition system, power system, which were designed to meet the requirement for each system. This report describes the details of the scaling and design features for the test facility

  12. New Sensors for the Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the selection strategy of what instrumentation is needed, and the program generated for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available to users of the ATR NSUF with data from irradiation tests using these sensors. In addition, progress is reported on current research efforts to provide users advanced methods for detecting temperature, fuel thermal conductivity, and changes in sample geometry

  13. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  14. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  15. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  16. Sharing of Rensselaer Polytechnic Institute Reactor Critical Facility (RCF)

    International Nuclear Information System (INIS)

    The RPI Reactor Critical Facility (RCF) operated successfully over the period fall 1994 - fall 1995. During this period, the RCF was used for Critical Reactor Laboratory spring 1995 (12 students); Reactor Operations Training fall 1994 (3 students); Reactor Operations Training spring 1995 (3 students); and Reactor Operations Training fall 1995 (3 students). Thirty-two Instrumentation and Measurement students used the RCF for one class for hands-on experiments with nuclear instruments. In addition, a total of nine credits of PhD thesis work were carried out at the RCF. This document constitutes the 1995 Report of the Rensselaer Polytechnic Institute's Reactor Critical Facility (RCF) to the USNRC, to the USDOE, and to RPI management

  17. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    The NNSA organized mainly in 1999 to complete the verification loop in core of the high flux experimental reactor with the 2000 kW fuel elements, the re-starting of China Pulsed Reactor, review and assessment on nuclear safety for the restarting of the Uranium-water critical Facility and treat the fracture event with the fuel tubes in the HWRR

  18. Design Guide for Category I reactors critical facilities

    International Nuclear Information System (INIS)

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

  19. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    International Nuclear Information System (INIS)

    The Department of Energy is currently engaged in a dual-track strategy to develop an accelerator and a commercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle'costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Department's purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work together 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay after 2005

  20. Models for transient analyses in advanced test reactors

    OpenAIRE

    Gabrielli, Fabrizio

    2011-01-01

    Several strategies are developed worldwide to respond to the world’s increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe whi...

  1. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

  2. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  3. Nuclear blenders: blended learning from Rensselaer's Reactor Critical Facility

    International Nuclear Information System (INIS)

    Rensselaer's senior level undergraduate nuclear engineering course 'Critical Reactor Laboratory' is highly regarded and much loved. If you can get in, that is. But now it's a required course, nuclear engineering enrollment is up, and others are knocking on our door to get in. How might one offer such a unique course to the masses, without losing the whole point of a laboratory experience? This presentation looks at the costs and benefits of the transition to a 'blended learning' mode -- the merging of traditional, face-to-face instruction and web-based instruction as a solution. As part of the presentation, the course and the facility will be highlighted by short excepts from the 50 minute movie 'Everything You Always Wanted to Know about Neutron Chain Reactions (but were afraid to ask)'.

  4. Enhanced in-pile instrumentation at the advanced test reactor

    International Nuclear Information System (INIS)

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

  5. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    International Nuclear Information System (INIS)

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

  6. The status of facilities at China Advanced Research Reactor

    International Nuclear Information System (INIS)

    A 60 MW research reactor, so called China Advanced Research Reactor (CARR,) was built in China Institute of Atomic Energy (CIAE), located in the southwest of Beijing and about 37 kilometers away from the central city. CARR is a tank-in-pool inverse neutron trap type reactor using D2O reflector, the designed optimal undisturbed thermal neutron flux is 8×1014 n⋅cm-2⋅s-1. A liquid D2 cold source will be equipped and the installation will be finished at the end of 2015. As a multipurpose research reactor, its main applications include neutron scattering, neutron activation analysis, isotope production, silicon doping, fuel element test, fundamental nuclear physics and so on. On March 13rd, 2012 CARR realized the 72 h stable operation with the full power. And the official operation license is expected to be issued at the beginning of next year. Cooperating with the internal and international users in the first phase ten instruments complete construction and are under commissioning, which are High Resolution Powder Diffractometer, High Intensity Powder Diffractometer, Residual Stress Diffractometer, Texture Diffractometer, Four Circle Diffractometer, Reflectometer, Small Angle Neutron Scattering, two Thermal Triple Axis Spectrometers and Isotope Separator On-Line instrument . In the second phase 7 instruments were approved and are under construction now. Although the operation license was not issued, the reactor was permitted to do the testing run several times and some results were obtained during the instrument commissioning.

  7. 10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), located in the ATR Complex of the Idaho National Laboratory (INL), was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. The ATR is fueled with high-enriched uranium (HEU) matrix (UAlx) in an aluminum sandwich plate cladding. The National Nuclear Security Administration Global Threat Reduction Initiative (GTRI) strategic mission includes efforts to reduce and protect vulnerable nuclear and radiological material at civilian sites around the world. Converting research reactors from using HEU to low-enriched uranium (LEU) was originally started in 1978 as the Reduced Enrichment for Research and Test Reactors (RERTR) Program under the U.S. Department of Energy (DOE) Office of Science. Within this strategic mission, GTRI has three goals that provide a comprehensive approach to achieving this mission: The first goal, the driver for the modification that is the subject of this determination, is to convert research reactors from using HEU to LEU. Thus the mission of the ATR LEU Fuel Conversion Project is to convert the ATR and Advanced Test Reactor Critical facility (ATRC) (two of the six U.S. High-Performance Research Reactors (HPRR)) to LEU fuel by 2017. The major modification criteria evaluation of the project pre-conceptual design identified several issues that lead to the conclusion that the project is a major modification.

  8. Reactor safety research program at Thai test facility

    International Nuclear Information System (INIS)

    Thermal-hydraulics, Hydrogen, Aerosol and Iodine (Thai) aims at providing experimental database for the verification and validation of Lumped Parameter (Lp) and Computational Fluid Dynamics (CFD) codes with 3-dimensional capabilities. Since its construction in 2000, Thai facility has been engaged in the field of reactor safety in the frame of various national (Thai I: 2000-2003, Thai II: 2003-2006, Thai III: 2006-2009, Thai IV: 2009-2012) and international programs (OECD-Thai: 2007-2009). Additionally, experimental data has been provided for several international standard problems (ISP 41, 46, 47 and 49) code validation exercises. Experiments performed in Thai facility cover a wide spectrum or reactor safety relevant issues by investigating separate and coupled-phenomenon experiments under design basis accident and severe-accident-typical scenarios. Experiments are performed in close co-operation with AREVA Erlangen and Grs Koln. Experimental configuration and the operating conditions in Thai vessel typical of those for PWR, BWR and High Temperature Gas Cooled Reactor can be produced thanks to its modular structure, appropriate feeding/generation devices for gases (H2, He, Steam, N2, etc.), Aerosol (inert and hygroscopic), Iodine Radiotracer, and advanced instrumentation. Experiments also cover investigation of passive safety systems, e.g. commercial Par for H2 mitigation in phenomenon orientated experiments to enhance the confidence in the performance of passive mitigation systems during severe accident scenarios and also to establish a common database accessible by a large research community to support further development and validation of the Lp and CFD codes with 3-dimensional capabilities. This paper summarizes experimental investigations made in Thai test facility to investigate issues related to the thermal-hydraulics, fission product (aerosol, iodine) transport and their interaction with containment walls (deposition, resuspension) and passive safety

  9. Models for transient analyses in advanced test reactors

    International Nuclear Information System (INIS)

    Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are applied

  10. Models for transient analyses in advanced test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gabrielli, Fabrizio

    2011-02-22

    Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are

  11. Reactor physics experiments with thorium based clusters in AHWR - critical facility

    International Nuclear Information System (INIS)

    AHWR - Critical Facility (AHWR - CF) is a 'zero power' reactor designed to carry out various reactor physics experiments for validation of AHWR design. A number of experiments have been carried out in standard and extended reference core of the reactor. In this paper, results of experiments with different Thorium based experimental clusters are presented. These experiments provided valuable data for validation of reactor physics design methodologies. (author)

  12. Data base of reactor physics experimental results in Kyoto University critical assembly experimental facilities

    International Nuclear Information System (INIS)

    The Kyoto University critical assembly experimental facilities belong to the Kyoto University Research Reactor Institute, and are the versatile critical assembly constructed for experimentally studying reactor physics and reactor engineering. The facilities are those for common utilization by universities in whole Japan. During more than ten years since the initial criticality in 1974, various experiments on reactor physics and reactor engineering have been carried out using many experimental facilities such as two solidmoderated cores, a light water-moderated core and a neutron generator. The kinds of the experiment carried out were diverse, and to find out the required data from them is very troublesome, accordingly it has become necessary to make a data base which can be processed by a computer with the data accumulated during the past more than ten years. The outline of the data base, the data base CAEX using personal computers, the data base supported by a large computer and so on are reported. (Kako, I.)

  13. Argonne to open new facility for advanced vehicle testing

    CERN Multimedia

    2002-01-01

    Argonne National Laboratory will open it's Advanced Powertrain Research Facility on Friday, Nov. 15. The facility is North America's only public testing facility for engines, fuel cells, electric drives and energy storage. State-of-the-art performance and emissions measurement equipment is available to support model development and technology validation (1 page).

  14. Construction of the two-phase critical flow test facility

    International Nuclear Information System (INIS)

    The two-phase critical test loop facility has been constructed in the KAERI engineering laboratory for the simulation of small break loss of coolant accident entrained with non-condensible gas of SMART. The test facility can operate at 12 MPa of pressure and 0 to 60 C of sub-cooling with 0.5 kg/s of non- condensible gas injection into break flow, and simulate up to 20 mm of pipe break. Main components of the test facility were arranged such that the pressure vessel containing coolant, a test section simulating break and a suppression tank inter-connected with pipings were installed vertically. As quick opening valve opens, high pressure/temperature coolant flows through the test section forming critical two-phase flow into the suppression tank. The pressure vessel was connected to two high pressure N2 gas tanks through a control valve to control pressure in the pressure vessel. Another N2 gas tank was also connected to the test section for the non-condensible gas injection. The test facility operation was performed on computers supported with PLC systems installed in the control room, and test data such as temperature, break flow rate, pressure drop across test section, gas injection flow rate were all together gathered in the data acquisition system for further data analysis. This test facility was classified as a safety related high pressure gas facility in law. Thus the loop design documentation was reviewed, and inspected during construction of the test loop by the regulatory body. And the regulatory body issued permission for the operation of the test facility

  15. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  16. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  17. A blueprint for GNEP advanced burner reactor startup fuel fabrication facility

    International Nuclear Information System (INIS)

    Research highlights: → This article discusses use of WG-plutonium as the startup fuel for Advanced Burner Reactor. → The presence of gallium in WG fuel may compromise the fuel integrity. → There is no facility exists to remove gallium from plutonium except at laboratory scale. → This article discusses the processes and issues associated with the gallium removal. → The article provides realistic scenario to all stack-holders involved in designing and operating ABR. - Abstract: The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu)-239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the

  18. Results of the critical experiments concerning OTTO loading at the critical HTR-test facility KAHTER

    International Nuclear Information System (INIS)

    Critical experiments concerning OTTO loading are described. In the KAHTER facility an OTTO loading has been simulated, therefore the original KAHTER assembly was reconstructed. The determination of critical masses and reactivity worths of control rods and of additional absorber rods in the top reflector and in the upper cavity was of main interest for comparison with reactor following calculations. Besides this, reaction rates in different energy regions were measured in the upper part of the core, in the cavity and top reflector. (orig.)

  19. Advanced neutron source corrosion test-loop facility

    International Nuclear Information System (INIS)

    The reference core for the advanced neutron source (ANS) will have a configuration similar to the present High-Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory: simply, an array of aluminum-alloy-clad fuel plates immersed in rapidly flowing water. The high thermal conductivity of the aluminum combined with the high heat transfer coefficient governing heat flow from the plate to the water keep the fuel inside the plates at an acceptable temperature. Unfortunately, the exposed of aluminum under these conditions leads to the formation of a thin layer of oxide (boehmite) that separates the fuel plates from the coolant water. The boehmite film has very poor thermal conductivity, and the heat flux that must cross this film can cause excessive heating of the fuel during the lifetime of the core. A test loop has been built to determine experimentally the rate of corrosion product formation on the aluminum cladding at the higher heat fluxes. Preliminary experiments have been completed that illustrate the capabilities of the loop system and the general behavior of an aluminum specimen exposed to large heat fluxes and coolant velocities. This same facility will be used for thermal-hydraulic testing; however, modifications will be necessary because of higher heat fluxes, temperatures, and pressures. Currently, the design is for light water testing; heavy water tests will be conducted in the future, which will require additional modifications

  20. EMERIS: an advanced information system for a materials testing reactor

    International Nuclear Information System (INIS)

    The basic features of the Materials Testing Reactor of IAE, Moscow (MR) Information System (EMERIS) are outlined. The purpose of the system is to support reactor and experimental test loop operators by a flexible, fully computerized and user-friendly tool for the aquisition, analysis, archivation and presentation of data obtained during operation of the experimental facility. High availability of EMERIS services is ensured by redundant hardware and software components, and by automatic configuration procedure. A novel software feature of the system is the automatic Disturbance Analysis package, which is aimed to discover primary causes of irregularities occurred in the technology. (author) 2 refs.; 2 figs

  1. Sharing of the RPI Reactor Critical Facility (RCF). Final summary report, January 1988--September 1995

    International Nuclear Information System (INIS)

    Rensselaer Polytechnic Institute (RPI) has participated for a number of years in Sharing of the Reactor Critical Facility (RCF) under the U.S. Department of Energy University Reactor Sharing Program. In September of each year a Sharing invitation is sent to 92 public and private high schools and to 74 colleges and universities within about a 3 hour drive to the RCF (Appendix B). Each year about 10 such educational institutions send groups to share the RCF

  2. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

  3. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  4. Modeling the critical hydrogen concentration in the AECL test reactor

    International Nuclear Information System (INIS)

    Hydrogen is added to a pressurized water reactor (PWR) to suppress radiolysis and maintain reducing conditions. The minimum hydrogen concentration needed to prevent radiolysis is referred to as the critical hydrogen concentration (CHC). The CHC was measured experimentally in the mid-1990s by Elliot and Stuart in a reactor loop at Atomic Energy of Canada (AECL), and was found to be approximately 0.5 scc/kg for typical PWR conditions. This value is well below industry-normal PWR operating levels near 40 scc/kg. Radiation chemistry models have also predicted a low CHC, even below the AECL experimental result. In the last few years some of the radiation chemical kinetic rate constants have been re-measured and G-values have been reassessed by Elliot and Bartels. These new data have been used in this work to revise the models and compare them with AECL experimental data. It is quite clear that the scavenging yields tabulated for high-LET radiolysis by Elliot and Bartels are not appropriate to use in the present context, where track-escape yields are needed to describe the homogeneous recombination kinetics in the mixed radiation field. In the absence of such data for high temperature PWR conditions, we have used the neutron G-values as fitting parameters. Even with this expedient, the model predicts at least a factor of two smaller CHC than was observed. We demonstrate that to recover the reported CHC result, the chemistry of ammonia impurity must be included. - Highlights: ► Hydrogen is added to nuclear reactor cooling loops to prevent radiolysis. ► Tests at AECL were carried out to determine the critical hydrogen concentration. ► Neutron radiolysis G-values need to be modified to understand the results. ► Ammonia impurity needs to be included for quantitative modeling.

  5. AREAL test facility for advanced accelerator and radiation source concepts

    Science.gov (United States)

    Tsakanov, V. M.; Amatuni, G. A.; Amirkhanyan, Z. G.; Aslyan, L. V.; Avagyan, V. Sh.; Danielyan, V. A.; Davtyan, H. D.; Dekhtiarov, V. S.; Gevorgyan, K. L.; Ghazaryan, N. G.; Grigoryan, B. A.; Grigoryan, A. H.; Hakobyan, L. S.; Haroutiunian, S. G.; Ivanyan, M. I.; Khachatryan, V. G.; Laziev, E. M.; Manukyan, P. S.; Margaryan, I. N.; Markosyan, T. M.; Martirosyan, N. V.; Mehrabyan, Sh. A.; Mkrtchyan, T. H.; Muradyan, L. Kh.; Nikogosyan, G. H.; Petrosyan, V. H.; Sahakyan, V. V.; Sargsyan, A. A.; Simonyan, A. S.; Toneyan, H. A.; Tsakanian, A. V.; Vardanyan, T. L.; Vardanyan, A. S.; Yeremyan, A. S.; Zakaryan, S. V.; Zanyan, G. S.

    2016-09-01

    Advanced Research Electron Accelerator Laboratory (AREAL) is a 50 MeV electron linear accelerator project with a laser driven RF gun being constructed at the CANDLE Synchrotron Research Institute. In addition to applications in life and materials sciences, the project aims as a test facility for advanced accelerator and radiation source concepts. In this paper, the AREAL RF photoinjector performance, the facility design considerations and its highlights in the fields of free electron laser, the study of new high frequency accelerating structures, the beam microbunching and wakefield acceleration concepts are presented.

  6. Education and training at the Rensselaer Polytechnic Institute reactor critical facility

    International Nuclear Information System (INIS)

    The Rensselaer Polytechnic Institute (RPI) Reactor Critical Facility (RCF) has provided hands-on education and training for RPI and other students for almost a quarter of a century. The RCF was built in the 1950s by the American Locomotive Company (ALCO) as a critical facility in which to carry out experiments in support of the Army Package power Reactor (APPR) program. A number of APPRs were built and operated. In the middle 1960s, ALCO went out of business and provided the facility to RPI. Since that time, RPI has operated the RCF primarily in a teaching mode in the nuclear engineering department, although limited amounts of reactor research, activation analysis, and reactivity assays have been carried out as well. Recently, a U.S. Department of Energy (DOE) upgrade program supported refueling the RCF with 4.81 wt% enriched UO2 high-density pellets clad in stainless steel rods. The use of these SPERT (F1) fuel rods in the RCF provided a cost-effective approach to conversion from high-enrichment bombgrade fuel to low-enrichment fuel. More important, however, is the fact that the new fuel is of current interest for light water power reactors with extended lifetime fuel. Thus, not only are critical reactor experiments being carried out on the fuel but, more importantly, the quality of the education and training has been enhanced

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

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States); Palmer, A.J.; Ingram, F.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Wiffen, F.W. [Dept. of Energy, Germantown, MD (United States). Office of Fusion Energy

    1998-09-01

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

  8. Experimental capabilities of the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    The TREAT facility was designed and built in the 1950s to provide a transient reactor for conducting safety experiments on reactor fuels. Throughout its almost 40-year history, it has proven to be a safe, reliable, and versatile facility, compiling a distinguished record of successful experiments. Several major improvements to the facility have been made, including an expansion of the building and of equipment handling capability, and enlargement of the access hole above the core, rearrangement of the reactor's control rods to provide more-uniform flux profiles, installation of improved reactor computer-control systems, a feedback system that safely allows real-time changes in power transients depending upon events occurring in the experiment, and several upgrades in the fast neutron hodoscope for improved experiment-fuel-motion diagnostics. The original TREAT fuel is still in use, however, since it appears to have no degradation from its many years of service

  9. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  10. Potential for new societal contributions from the advanced test reactor

    International Nuclear Information System (INIS)

    The mission of the Advanced Test Reactor (ATR) at Idaho National Engineering Laboratory is to study the effects of intense radiation on materials and fuels and to produce radioisotopes for the U.S. Department of Energy (DOE) for government and commercial applications. The purpose of this paper is to explore the potential benefits to society from these available neutrons. The ATR is a 250-MW(thermal) light water reactor highly enriched uranium in plate-type fuel. The ATR uses a combination of hafnium control drums and shim rods to adjust power and hold flux distortion to a minimum. The different quadrants of the ATR can be operated at different power levels to meet a variety of mission requirements. Irradiation positions are available at various locations throughout the core and beryllium reflector. A summary of the flux levels at various ATR reflector and loop positions. is given. These fluxes are maintained with a relatively constant axial flux profile throughout cycles that last 35 to 42 days. These neutrons can be used for testing and irradiation programs that support commercial reactor license extension, advanced fuel development, materials effects studies, failure cause/effect studies, coolant chemistry evaluations, prototype testing programs, isotope production, and basic research. Radioisotope production falls into three categories: medical, industrial, and research. In summary, the ATR is a unique, high-power test reactor capable of supporting the current DOE mission and producing radioisotopes. Space available for radioisotope production and fuels or materials testing will increase by 44% in 1994, improving DOE's ability to support national needs in health care, industry, and research

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

    2011-09-01

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

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

    Science.gov (United States)

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

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. PMID:21399407

  14. Replacement of the Advanced Test Reactor control room

    International Nuclear Information System (INIS)

    The control room for the Advanced Test Reactor has been replaced to provide modern equipment utilizing current standards and meeting the current human factors requirements. The control room was designed in the early 1960 era and had not been significantly upgraded since the initial installation. The replacement did not change any of the safety circuits or equipment but did result in replacement of some of the recorders that display information from the safety systems. The replacement was completed in concert with the replacement of the control room simulator which provided important feedback on the design. The design successfully incorporates computer-based systems into the display of the plant variables. This improved design provides the operator with more information in a more usable form than was provided by the original design. The replacement was successfully completed within the scheduled time thereby minimizing the down time for the reactor

  15. Preliminary safety evaluation of the advanced burner test reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, F. E.; Fanning, T. H.; Cahalan, J. E.; Nuclear Engineering Division

    2006-09-15

    Results of a preliminary safety evaluation of the Advanced Burner Test Reactor (ABTR) pre-conceptual design are reported. The ABTR safety design approach is described. Traditional defense-in-depth design features are supplemented with passive safety performance characteristics that include natural circulation emergency decay heat removal and reactor power reduction by inherent reactivity feedbacks in accidents. ABTR safety performance in design-basis and beyond-design-basis accident sequences is estimated based on analyses. Modeling assumptions and input data for safety analyses are presented. Analysis results for simulation of simultaneous loss of coolant pumping power and normal heat rejection are presented and discussed, both for the case with reactor scram and the case without reactor scram. The analysis results indicate that the ABTR pre-conceptual design is capable of undergoing bounding design-basis and beyond-design-basis accidents without fuel cladding failures. The first line of defense for protection of the public against release of radioactivity in accidents remains intact with significant margin. A comparison and evaluation of general safety design criteria for the ABTR conceptual design phase are presented in an appendix. A second appendix presents SASSYS-1 computer code capabilities and modeling enhancements implemented for ABTR analyses.

  16. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  17. Temperature controlled material irradiation in the advanced test reactor

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) is located at the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho, USA and is owned and regulated by the U.S. Department of Energy (US DOE). The ATR is operated for the US DOE by Lockheed Martin Idaho Technologies. In recent years, prime irradiation space in the ATR has been made available for use by customers having irradiation service needs in addition to the reactor's principal user, the U.S. Naval Nuclear Propulsion Program. To enhance the reactor's capabilities, the US DOE has initiated the development of an Irradiation Test Vehicle (ITV) capable of providing neutron spectral tailoring and temperature control for up to 28 experiments. The ATR-ITV will have the flexibility to simultaneously support a variety of experiments requiring fast, thermal or mixed spectrum neutron environments. Temperature control is accomplished by varying the thermal conductivity across a gas gap established between the experiment specimen capsule wall and the experiment 'in-pile tube (IPT)' inside diameter. Thermal conductivity is adjusted by alternating the control gas mixture ratio of two gases with different thermal conductivities

  18. Test on the reactor with the intelligent extrapolation criticality device for physical startup experiment

    International Nuclear Information System (INIS)

    The Intelligent Extrapolation Criticality Device is used for automatic counting and automatic extrapolation during the criticality experiment on the reactor. Test must be performed on the zero-power reactor or other reactor before the Device is used. The paper describes the test situation and test results of the Device on the zero-power reactor. The test results show that the Device has the function of automatic counting and automatic extrapolation, the deviation of the extrapolation data is small, and it can satisfy the requirements of physical startup on the reactor. (author)

  19. The ''CAMERA'' test facility in the OSIRIS reactor

    International Nuclear Information System (INIS)

    CAMERA is an irradiation installation conceived to measure under neutronic flux and continuously the dimension variations of a fuel pencil of PWR reactors. The device, set in the periphery of the OSIRIS reactor, can receive new, preirradiated or reconstituted pencils. The principles of measurements is explained. Then, a brief description of the installation is given: in-pile part; out-of-pile part; connections. The technical characteristics of the installation are presented. A first qualification test of the installation under flux has been carried out at the end of the first semester 1984 in the OSIRIS reactor

  20. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pope, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morrell, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jamison, R. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nef, E. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nigg, D. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  1. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  2. The advanced containment experiments (ACE) radioiodine test facility experimental program

    International Nuclear Information System (INIS)

    Results of the Advanced Containment Experiments (ACE) Radioiodine Test Facility (RTF) program are reported. This study consisted of four intermediate-scale experiments that investigated the effects of radiation, pH, surfaces and initial iodine speciation on iodine behaviour. The tests revealed that, in high radiation fields, the long-term volatility of iodine is independent of the initial iodine speciation (CsI, I2, CH3I). This is presumably because radiolytic reactions inter-convert aqueous iodine species; I- was the predominant aqueous iodine species after an absorbed dose of about 30-40 kGy. Tests at pH 9 and 5.5 demonstrated that iodine volatility increased significantly with decreasing pH. In addition, this study demonstrated that containment surfaces can play an important role in determining iodine volatility, gas and aqueous phase iodine speciation, and surface adsorption. In summary: The ACE/RTF experiments have demonstrated the importance of several factors on iodine behaviour within containment under reactor accident conditions. One of the most important factors was radiation. Without radiation, the volatility of iodine was dependent on the initial speciation of iodine, presumably because inter-conversion of iodine species by non-radiolytic reactions is relatively slow. In contrast, in the presence of radiation, the long-term volatility of iodine was independent of initial speciation. This is attributed to aqueous phase radiolytic reactions that result in rapid inter-conversion of iodine species. Iodine volatility was shown to increase significantly with decreasing pH. However, changing the pH from acidic to alkaline conditions did not result in rapid decreases in iodine volatility. This may have been due to desorption of volatile iodine species from surfaces, in the case of stainless steel, and the influence of organics in the epoxy tests. Surfaces were shown to influence iodine volatility and speciation. Higher gas phase iodine concentrations were

  3. Conceptual design study advanced concepts test (ACT) facility

    Energy Technology Data Exchange (ETDEWEB)

    Zaloudek, F.R.

    1978-09-01

    The Advanced Concepts Test (ACT) Project is part of program for developing improved power plant dry cooling systems in which ammonia is used as a heat transfer fluid between the power plant and the heat rejection tower. The test facility will be designed to condense 60,000 lb/hr of exhaust steam from the No. 1 turbine in the Kern Power Plant at Bakersfield, CA, transport the heat of condensation from the condenser to the cooling tower by an ammonia phase-change heat transport system, and dissipate this heat to the environs by a dry/wet deluge tower. The design and construction of the test facility will be the responsibility of the Electric Power Research Institute. The DOE, UCC/Linde, and the Pacific Northwest Laboratories will be involved in other phases of the project. The planned test facilities, its structures, mechanical and electrical equipment, control systems, codes and standards, decommissioning requirements, safety and environmental aspects, and energy impact are described. Six appendices of related information are included. (LCL)

  4. Modelling activities of experimental facilities related to advanced reactors. Considerations on 1D/3D issues

    International Nuclear Information System (INIS)

    The state of art of modelling activities related to integral experimental facilities of advanced passive reactors show to date important open items. The main advantage of using 1D plant codes is the capability of simulating the full interaction between components traditionally correctly modelled (condensers, heat exchangers, pipes and vessels) and other components for which codes are not 100% suitable (pools and containments). Polytechnical University of Catalonia (UPC) and Polytechnical University of Valencia (UPV) cooperated with other European research organizations in the 'Technology Enhancement for Passive Safety Systems' (TEPSS) project, within the European Fourth Framework Programme. It was a task of both Universities to supply analytical support of PANDA tests. The paper deals with the 1D/3D discussion in the framework of modelling activities related to integral passive facilities like PANDA. It starts choosing reference tests among those corresponding to our participation in TEPSS project. The discrepancies observed in a 1D simulation of the selected tests will be shown and analyzed. An evaluation of how the 3D version can lead to a better agreement with data will be included. Disadvantages of 3D codes will be shown too. Combining the use of different codes, and considering analyst criteria, will make possible to establish suitable recommendations from both engineering and scientific point of view. (author)

  5. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  6. Communication of Advanced Test Reactor probabilistic risk assessment results

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) probabilistic risk assessment (PRA) Level 1 results were efficiently communicated in two reports following the completion of revision 1 of the ATR PRA. As the ATR PRA including external events fills four large volume, it was considered impractical to expect all of the individuals at ATR who could benefit from the information to read the entire PRA. Even though many ATR personnel received training in PRA methodology and were involved in some aspects of the PRA, another hinderance to effective communication of the PRA results is that the PRA was written and organized to meet the requirements of practitioners and reviewers who are well-versed in PRA methods. Therefore, two PRA summary reports, an ATR risk summary report and an ATR functional group summary report, were written to communicate the ATR PRA results and insights to interested ATR personnel

  7. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2007-09-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  8. Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

    International Nuclear Information System (INIS)

    This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

  9. IPPE critical facilities and their research programs

    International Nuclear Information System (INIS)

    The 40th anniversary of BFS zero power fast critical facilities family took place in 2001. An extensive neutron physics research program for justification of fast sodium-cooled reactors core physics has been carried out on them. Advanced reactors core physics research is fulfilled today to solve both traditional and non-traditional tasks of nuclear power industry

  10. A review of two recent occurrences at the Advanced Test Reactor involving subcontractor activities

    International Nuclear Information System (INIS)

    This report documents the results of a brief, unofficial investigation into two incidents at the Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR) facility, reported on October 25 and 31, 1997. The first event was an unanticipated breach of confinement. The second involved reactor operation with an inoperable seismic scram subsystem, violating the reactor's Technical Specifications. These two incidents have been found to be unrelated. A third event that occurred on December 16, 1996, is also discussed because of its similarities to the first event listed above. Both of these incidents were unanticipated breaches of confinement, and both involved the work of construction subcontractor personnel. The cause for the subcontractor related occurrences is a work control process that fails to effectively interface with LMITCO management. ATR Construction Project managers work sufficient close with construction subcontractor personnel to understand planned day-to-day activities. They also have sufficient training and understanding of reactor operations to ensure adherence to applicable administrative requirements. However, they may not be sufficiently involved in the work authorization and control process to bridge an apparent communications gap between subcontractor employees and Facility Operations/functional support personnel for work inside the reactor facility. The cause for the inoperable seismic scram switch (resulting from a disconnected lead) is still under investigation. It does not appear to be subcontractor related

  11. Authentication system for the JAERI Fast Critical Facility Advanced Containment and Surveillance System

    International Nuclear Information System (INIS)

    In a joint effort conducted by Sandia National Laboratories, the International Atomic Energy Agency (IAEA), and the Japan Atomic Energy Research Institute (JAERI), an authentication system has been installed at the Fast Critical Assembly (FCA) facility in Tokai-mura, Japan. The purpose of this authentication system is to provide the IAEA with an independent means of authenticating the operator-provided Advanced Containment and Survellance (AC/S) system already in place at the facility. Authentication Controllers were installed at the AC/S Portal Monkor and Penetration Monitor to collect data and to randomly test sensor functions between IAEA inspections. During each inspection the authentication data is collected with an Inspector's portable computer and printed for comparison to the data recorded by the AC/S system. Installation of the authentication equipment took place in November 1991 and a three-month field test began in December 1991. This paper will describe the authentication system, the operator interface, and the preliminary results of the field tests

  12. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  13. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  14. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    International Nuclear Information System (INIS)

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection

  15. Operating the Advanced Test Reactor in today's economic and regulatory environment

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory, is the US Department of Energy's largest and most versatile test reactor. Base programs at ATR are planned well into the 21st century. The ATR and support facilities along with an overview of current programs will be reviewed, but the main focus of the presentation will be on the impact that today's economic and regulatory concerns have had on the operation of this test reactor. Today's economic and regulatory concerns have demanded more work be completed at lower cost while increasing the margin of safety. By the beginning of the 1990 s, federal budgets for research generally and particularly for nuclear research had decreased dramatically. Many national needs continued to require testing in the ATR; but demanded lower cost, increased efficiency, improved performance, and an increased margin of safety. At the same time budgets were decreasing, there was an increase in regulatory compliance activity. The new standards imposed higher margins of safety. The new era of greater openness and higher safety standards complemented research demands to work safer, smarter and more efficiently. Several changes were made at the ATR to meet the demands of the sponsors and public. Such changes included some workforce reductions, securing additional program sponsors, upgrading some facilities, dismantling other facilities, and implementing new safety programs. (author)

  16. Lead Coolant Test Facility Development Workshop

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz

    2005-06-01

    A workshop was held at the Idaho National Laboratory on May 25, 2005, to discuss the development of a next generation lead or lead-alloy coolant test facility. Attendees included representatives from the Generation IV lead-cooled fast reactor (LFR) program, Advanced Fuel Cycle Initiative, and several universities. Several participants gave presentations on coolant technology, existing experimental facilities for lead and lead-alloy research, the current LFR design concept, and a design by Argonne National Laboratory for an integral heavy liquid metal test facility. Discussions were focused on the critical research and development requirements for deployment of an LFR demonstration test reactor, the experimental scope of the proposed coolant test facility, a review of the Argonne National Laboratory test facility design, and a brief assessment of the necessary path forward and schedule for the initial stages of this development project. This report provides a summary of the presentations and roundtable discussions.

  17. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2006-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  18. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The United States Department of Energy's Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation

  19. Test Results of Reactor Coolant System Natural Circulation using the SMART-ITL Facility

    International Nuclear Information System (INIS)

    In this paper, the Sequence Of Event (SOE) and test conditions of RCS natural circulation test using SMART-ITL are presented, and the major measuring parameters and the test results will be introduced. In this test, the steady state operation satisfied the initial condition of the prescribed test procedure and the boundary conditions were properly simulated. After the RCPs stop, the RCS natural circulation flow was generated by heating in the core region and cooling in the SG heat exchanger region, and the major thermalhydraulic parameters reached at a stable condition. Through this experiment, it has been validated that the SMART-ITL facility can adequately simulate the RCS natural circulation behavior. In addition, it is expected that the experimental data can be used for the code assessment of the TASS/SMR-S code and experiences from this test can be utilized to the subsequent SBLOCA simulation test. SMART (System-integrated Modular Advanced Reactor) is an integral type reactor which major primary components such as the steam generator, the pressurizer, and the reactor coolant pump are installed inside one single reactor vessel and connecting primary pipes are removed. The TASS/SMR-S code is used to perform the performance and safety analysis of the SMART. To evaluate the capability of TASS/SMR-S code on the natural circulation and accident scenarios such as Small-Break Loss of Coolant Accident (SBLOCA) for predicting the thermal-hydraulic phenomena in steady state and transient operation, it is essential to perform a series of validation tests

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

    International Nuclear Information System (INIS)

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

  1. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  2. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

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

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  5. Design, development, manufacturing, testing and commissioning of instrumentation for critical facility

    International Nuclear Information System (INIS)

    The Critical Facility (CF) is a low power research reactor. Neutronic and Process Instrumentation are important systems those ensure safety and control of the reactor. The Control and Instrumentation systems for the Critical facility have been designed, developed and commissioned by RRMD. Most of the nuclear instrumentation used here is originally designed and developed jointly by Electronics Division and RRMD during ZAC project for conversion of the earlier vacuum tube based instrumentation to the solid state circuits during 1984-85 for Zerlina, Apsara and Cirus and are in operation at Cirus and Apsara since commissioning in 1984-85. The Process Instrumentation is provided for monitoring important process parameters such as level, flow, temperature and pressure and generating trips, alarms and interlocks necessary for smooth operation of the systems and safety of personnel and equipment. The paper will provide technical information on the reactor instrumentation for both nuclear and process systems, used in Critical Facility, the associated trip and alarms logics, and their integration into the reactor monitoring control and protection system, information systems developed for Critical Facility. (author)

  6. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program

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

    International Nuclear Information System (INIS)

    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 been 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.

  8. Reactor fault simulation at the closure of the Windscale advanced gas-cooled reactor: analysis of reactor transient tests

    International Nuclear Information System (INIS)

    The testing of fault transient analysis methods by direct simulation of fault sequences on a commercial reactor is clearly excluded on safety and economic grounds. The closure of the Windscale prototype advanced gas-cooled reactor (WAGR) therefore offered a unique opportunity to test fault study methods under extreme conditions relatively unfettered by economic constraints, although subject to appropriate safety regulations. One aspect of these important experiments was a series of reactor transient tests. The objective of these reactor transients was to increase confidence in the fault study computer models used for commercial AGR safety assessment by extending their range of validation to cover large amplitude and fast transients in temperature, power and flow, relevant to CAGR faults, and well beyond the conditions achievable experimentally on commercial reactors. A large number of tests have now been simulated with the fault study code KINAGRAX. Agreement with measurement is very good and sensitivity studies show that such discrepancies as exist may be due largely to input data errors. It is concluded that KINAGRAX is able to predict steady state conditions and transient amplitudes in both power and temperature to within a few percent. (author)

  9. GERDA test facility for pressurized water reactors with straight tube steam generators

    International Nuclear Information System (INIS)

    A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.)

  10. Underwater plasma arc cutting of in-reactor tube of In-Pile Creep Test Facility

    International Nuclear Information System (INIS)

    The in-reactor tube of the In-Pile Creep Facility had been irradiated periodically for over 6 years in the Japan Materials Testing Reactor (JMTR) up to the end of 1978 under an operating condition of high temperature and high pressure identical to that of the Prototype Advanced Thermal Reactor, FUGEN, to gain the basic data for estimating the amount of creep which would occur in the pressure tubes of FUGEN. Following the removal of the in-reactor tube out of the JMTR, the test sections in the tube which were to be subjected to post irradiation examination were cut out. Underwater plasma arc cutting was employed to prevent the spread of contamination to the work area, to confine the heat affected zone in the test pieces to a minimum and to simplify disposal of the unneeded portions of the pressure tube. Setup of the cutting machine, cutting operations, radiological conditions during cutting of the highly radioactive portion of the tube and disassembly of the cutting equipment are described. In addition a brief description of the underwater plasma arc cutting machine is presented. The hot-cutting operations were done remotely to control personal exposure. The containment envelope prevented the spread of contamination to the environment and radioactive particles deposited on the cutting machine were removed without any difficulties. External exposure received by cutting personnel was small. Internal radionuclide deposit examinations were conducted, determining no crew member inhaled radioactive substances. Contamination spreads to the work area were minimal and release of radionuclide was well controlled. (author)

  11. Design considerations of the irradiation test vehicle for the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  12. Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

  13. Reactor coolant cleanup facility

    International Nuclear Information System (INIS)

    A depressurization device is disposed in pipelines upstream of recycling pumps of a reactor coolant cleanup facility to reduce a pressure between the pressurization device and the recycling pump at the downstream, thereby enabling high pressure coolant injection from other systems by way of the recycling pumps. Upon emergency, the recycling pumps of the coolant cleanup facility can be used in common to an emergency reactor core cooling facility and a reactor shutdown facility. Since existent pumps of the emergency reactor core cooling facility and the reactor shutdown facility which are usually in a stand-by state can be removed, operation confirmation test and maintenance for equipments in both of facilities can be saved, so that maintenance and reliability of the plant are improved and burdens on operators can also be mitigated. Moreover, low pressure design can be adopted for a non-regenerative heat exchanger and recycling coolant pumps, which enables to improve the reliability and economical property due to reduction of possibility of leakage. (N.H.)

  14. Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA Critical Facility

    International Nuclear Information System (INIS)

    The IAEA has facilitated an extensive programme that addresses the technical development of advanced gas cooled reactor technology. Included in this programme is the coordinated research project (CRP) on Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance, which is the focus of this TECDOC. This CRP was established to foster the sharing of research and associated technical information among participating Member States in the ongoing development of the HTGR as a future source of nuclear energy. Within it, computer codes and models were verified through actual test results from operating reactor facilities. The work carried out in the CRP involved both computational and experimental analysis at various facilities in IAEA Member States with a view to verifying computer codes and methods in particular, and to evaluating the performance of HTGRs in general. The IAEA is grateful to China, the Russian Federation and South Africa for providing their facilities and benchmark programmes in support of this CRP.

  15. Renovation of CPF (Chemical Processing Facility) for Development of Advanced Fast Reactor Fuel Cycle System

    International Nuclear Information System (INIS)

    CPF (Chemical Processing Facility) was constructed at Nuclear Fuel Cycle Engineering Laboratories of JAEA (Japan Atomic Energy Agency) in 1980 as a basic research field where spent fuel pins from fast reactor (FR) and high level liquid waste can be dealt with. The renovation consists of remodeling of the CA-3 cell and the laboratory A, installation of globe boxes, hoods and analytical equipments to the laboratory C and the analytical laboratory. Also maintenance equipments in the CA-5 cell which had been out of order were repaired. The CA-3 cell is the main cell in which important equipments such as a dissolver, a clarifier and extractors are installed for carrying out the hot test using the irradiated FR fuel. Since the CPF had specialized originally in the research function for the Purex process, it was desired to execute the research and development of such new, various reprocessing processes. Formerly, equipments were arranged in wide space and connected with not only each other but also with utility supply system mainly by fixed stainless steel pipes. It caused shortage of operation space in flexibility for basic experimental study. Old equipments in the CA-3 cell including vessels and pipes were removed after successful decontamination, and new equipments were installed conformably to the new design. For the purpose of easy installation and rearranging the experimental equipments, equipments are basically connected by flexible pipes. Since dissolver is able to be easily replaced, various dissolution experiments is conducted. Insoluble residue generated by dissolution of spent fuel is clarified by centrifugal. This small apparatus is effective to space-saving. Mini mixer settlers or centrifugal contactors are put on to the prescribed limited space in front of the backside wall. Fresh reagents such as solvent, scrubbing and stripping solution are continuously fed from the laboratory A to the extractor by the reagent supply system with semi-automatic observation

  16. Safety review and assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    More operational events were occurred at various research reactors in 1995. The NNSA and its regional offices conducted careful investigation and strict regulation. In order to analyze comprehensively the safety situation of inservice research reactors and find same countermeasures the NNSA convened a meeting of the safety regulation on research reactors and a meeting for change experience of the safety regulation on research reactors that were participated in by the operating organizations in 1995. A lot of work has been done in the respects of propagation of regulations on nuclear safety, education of nuclear safety culture, the investigation and treatment of operational events, the reexamine of operation documents, the implementation of rectifying items on nuclear safety, the daily inspection and routine inspection on nuclear safety and the studying on the extending service life of research reactors etc

  17. Refurbishment status on reactor facilities of JMTR

    International Nuclear Information System (INIS)

    The JMTR (Japan Materials Testing Reactor), a light-water-cooling tank-type reactor with a 50 MW thermal power, was shutdown in August 2006. The reactor facilities are to be refurbished during four years from the beginning of FY 2007, and the renewed JMTR will restart from FY 2011. In advance of the reactor refurbishment, equipments on reactor facilities to be renewed and to be continuously used were selected from a viewpoint of ensuring safety, improvement of operating availability, etc. The selected equipments to be renewed were the reactor instrument and control system, cooling system, radioactive waste facility, power supply system, boiler, etc. This report describes the basic idea on selection of the renewal facilities and schedule of refurbishment work. (author)

  18. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-09-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  19. Sodium natural convection testing in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility

    International Nuclear Information System (INIS)

    A comparison is made between experimental data and analytical results for a single-phase natural convection test in an experimental sodium loop. The test was conducted in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale high temperature sodium loop at the Oak Ridge National Laboratory (ORNL), used for thermal-hydraulic testing of simulated Liquid Metal Fast Breeder Reactor (LMFBR) subassemblies at normal and off-normal operating conditions. Electrical heating in the 19-pin assembly during the test was typical of decay heat levels. The test chosen for analysis in this paper was one of seven natural convection runs conducted in the facility. In this test the bypass line was open to simulate a parallel heated assembly and the test was begun with a pump coastdown from a small initial forced flow

  20. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    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. 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.

  1. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; J. E. Daw

    2011-03-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership 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, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this 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. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  2. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership 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, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this 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. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  3. Severe-fuel-damage experiments in the Canadian in-reactor Blowdown Test Facility

    International Nuclear Information System (INIS)

    The Blowdown Test Facility consists of an instrumented in-reactor irradiation site plus an out-reactor piping system. These are used to irradiate CANDU fuel under conditions representative of a loss-of-coolant accident (LOCA) or LOCA with Loss-Of-Emergency-Core-Cooling (LOECC) in order to study fuel performance, fission-product release from the fuel and the transport of fission products through the piping system. An overview of the facility and the experimental program is given in this paper. (author)

  4. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  5. Design and fabrication of sodium test facility for fast breeder reactor

    International Nuclear Information System (INIS)

    The purpose of the promotion policy for energy research and development base construction plan (priority facility) of the Japanese government in FY2009 is 'to construct in Tsuruga City the research and development base for plant operation technology for the practical use of fast breeder reactor where researchers in and out of Japan gather, and to contribute to the development and revitalization of the region as the base with international characteristics.' In conformity to this purpose, the Japan Atomic Energy Agency built 'sodium engineering research facilities' in Tsuruga. This paper describes the design, fabrication, and installation of interior equipment that were carried out by Kawasaki Heavy Industries. 'Sodium engineering research facilities' are the test and research facilities to conduct research and development related to sodium, while reflecting the experiences of operation and maintenance of 'Monju,' which aims at the commercialization of fast reactor. The facilities specialize in the handling technology of sodium to meet the needs in and out of Japan, and were completed in June 2015. The facilities consist of six units including tank-loop test equipment, mini-loop test equipment, sodium purification and supply equipment, etc. For the tank-loop test equipment, a sodium transfer test of about 5.5 tons, and a subsequent comprehensive function test using sodium are scheduled. (A.O.)

  6. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    International Nuclear Information System (INIS)

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition

  7. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.

  8. Improved Computational Neutronics Methods And Validation Protocols For The Advanced Test Reactor

    International Nuclear Information System (INIS)

    The Idaho National Laboratory (INL) is in the process of modernizing the various reactor physics modeling and simulation tools used to support operation and safety assurance of the Advanced Test Reactor (ATR). Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009 was successfully completed during 2011. This demonstration supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR fuel cycle management process beginning in 2012. On the experimental side of the project, new hardware was fabricated, measurement protocols were finalized, and the first four of six planned physics code validation experiments based on neutron activation spectrometry were conducted at the ATRC facility. Data analysis for the first three experiments, focused on characterization of the neutron spectrum in one of the ATR flux traps, has been completed. The six experiments will ultimately form the basis for a flexible, easily-repeatable ATR physics code validation protocol that is consistent with applicable ASTM standards.

  9. The operation of the Tokamak Fusion Test Reactor Tritium Facility

    Energy Technology Data Exchange (ETDEWEB)

    Gentile, C.A.; LaMarche, P.H. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Anderson, J.L. [Los Alamos National Lab., NM (United States)

    1995-07-01

    The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During `` time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems.

  10. Evaluation of the advanced mixed-oxide fuel test FO-2 irradiated in the FFTF [Fast Flux Test Facility

    International Nuclear Information System (INIS)

    The advanced mixed-oxide (UO2-PuO2) test assembly, FO-2, irradiated in the Fast Flux Test Facility (FFTF) is undergoing postirradiation examination. This is one of the first FFTF tests examined that used the advanced ferrite-martensite alloy, HT9, which is highly resistant to irradiation swelling. The FO-2 includes the first annular fueled pins irradiated in FFTF to undergo destructive examination. The FO-2 is a lead assembly for the ongoing FFTF Core Demonstration Experiment (CDE) and was designed to evaluate the effects of fuel design variables, such as pellet density, smeared density, and fuel form (annular or solid fuel), on advanced pin performance. The assembly contains a total of 169 fuel pins of 12 different types. Two L (annular) fuel pins, GF02L04 (FFTF and transient tested) and GF02L09 (FFTF only), were destructively examined. Evaluation of the FO-2 fuel pins and assembly shows the excellent and predictable performance of the mixed-oxide fuels with HT9 structural material. This, combined with the robust behavior of the pins in transient tests, and the continued excellent performance of the CDE indicate this is a superior fuel system for liquid-metal reactors. It offers greatly reduced deformation during irradiation, while maintaining good operating characteristics

  11. Advances in sodium technology, testing and diagnostics of fast reactors

    International Nuclear Information System (INIS)

    The collection contains a selection of 29 papers from three international specialists' meetings: the CMEA conference ''Control and measuring instruments and diagnostic systems of fast reactors'' held in the GDR in April 1983; the IAEA conference on nuclear power experience held in Austria in September 1982; and the conference ''Problems of technology and corrosion in sodium coolant and protective gas'' held in the GDR in April 1977. Three papers on operating experience with Soviet fast reactors and their safety have a general character; they are followed up by three papers on sodium technology. Five papers deal with the diagnostics of fast sodium cooled reactors and nine papers are devoted to the diagnostics of steam generators. Eight papers relate to detectors for the diagnostics of fast reactors. Safety regulations for work with alkali metals are added. (A.K.)

  12. Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project

    Energy Technology Data Exchange (ETDEWEB)

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  13. Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project

    Energy Technology Data Exchange (ETDEWEB)

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  14. Verification of core-fuel inventory of a fast critical facility by monitoring reactor physics parameters

    International Nuclear Information System (INIS)

    On the safeguards problem, a technical feasibility was studied for experimentally verifying core-fuel inventory of a fast critical facility. The FCA Assembly VIII-1 with plutonium-fueled test zone was used for this purpose. Six loading patterns were chosen in the verification experiment to simulate the diversion of Pu-fuel from the core. The Pu-fuel removed from the core was about 3.5 -- 5.8 kg. Verification techniques are based on the monitoring of small changes in fission rates and #betta#/l caused by the diversion of some amount of Pu-fuel. The fission rates were measured by a fission chamber technique with a hundred 239Pu fission chambers located in the core region and multi-chamber scanning electronics, while #betta#/l values were measured by power noise analysis with two herium-3 chambers. The verification experiment indicates that the fission rates and #betta#/l monitor well follow the quantities of plutonium removed from the core. It is concluded that the verification of core-fuel inventory is feasible by using the present monitoring method. (author)

  15. Nuclear Safety Research Reactor (NSRR) as a facility for reactor safety research and its modification for the future test plan

    International Nuclear Information System (INIS)

    The NSRR is a modified TRIGA-ACPR (annular core pulse reactor), and attained the initial criticality in May, 1975. It was built for studying reactor fuel behavior under a reactivity-initiated accident condition. The reactor is installed in a pool of 3.6 m width, 4.5 m length and 9 m depth, and water above the reactor core serves as a radiation shield. The reactor core contains 149 driver fuel rods, 6 regulating rods, 2 safety rods and 3 transient rods. An arbitrary reactivity up to 4.67 $ can be set up almost instantaneously in the reactor core. The pulse power generation is terminated by the large negative reactivity induced by prompt temperature feedback without inserting the control rods. This is brought about by an excellent property of the driver fuel which contains 12 wt.% U-ZrH enriched to 20 wt.% U-235. As a unique feature, the NSRR is equipped with a big experimental cavity through the center of the reactor core. It has the diameter of 220 mm, and is called loading tube. It is branched into a vertical loading tube and an offset loading tube. The characteristics of the pulse operation in the NSRR, the outline of fuel irradiation experiment, the future test plan and the modification of the NSRR are described. (Kako, I.)

  16. Evaluation of the advanced mixed oxide fuel test FO-2 irradiated in Fast Flux Test Facility

    International Nuclear Information System (INIS)

    The advanced mixed-oxide (UO2-PuO2) test assembly, FO-2, irradiated in the Fast Flux Test Facility (FFTF), is undergoing postirradiation examination (PIE). This is one of the first FFTF tests examined that used the advanced ferrite-martensite alloy, HT9, which is highly resistant to irradiation swelling. The FO-2 includes the first annular fueled pins irradiated in FFTF to undergo destructive examination. The FO-2 is a lead assembly for the ongoing FFTF Core Demonstration Experiment (CDE) (Leggett and Omberg 1987) and was designed to evaluate the effects of fuel design variables, such as pellet density, smeared density, and fuel form (annular or solid fuel), on advanced pin performance. The assembly contains a total of 169 fuel pins of twelve different types. The test was irradiated for 312 equivalent full power days (EFPD) in FFTF. It had a peak pin power of 13.7 kW/ft and reached a peak burnup of 65.2 MWd/kgM with a peak fast fluence of 9.9 /times/ 1022 n/cm2 (E > 0.1 MeV). This document discusses the test and its results. 6 refs., 19 figs., 4 tabs

  17. Scoping analysis of the Advanced Test Reactor using SN2ND

    Energy Technology Data Exchange (ETDEWEB)

    Wolters, E.; Smith, M. (NE NEAMS PROGRAM); ( SC)

    2012-07-26

    A detailed set of calculations was carried out for the Advanced Test Reactor (ATR) using the SN2ND solver of the UNIC code which is part of the SHARP multi-physics code being developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in DOE-NE. The primary motivation of this work is to assess whether high fidelity deterministic transport codes can tackle coupled dynamics simulations of the ATR. The successful use of such codes in a coupled dynamics simulation can impact what experiments are performed and what power levels are permitted during those experiments at the ATR. The advantages of the SN2ND solver over comparable neutronics tools are its superior parallel performance and demonstrated accuracy on large scale homogeneous and heterogeneous reactor geometries. However, it should be noted that virtually no effort from this project was spent constructing a proper cross section generation methodology for the ATR usable in the SN2ND solver. While attempts were made to use cross section data derived from SCALE, the minimal number of compositional cross section sets were generated to be consistent with the reference Monte Carlo input specification. The accuracy of any deterministic transport solver is impacted by such an approach and clearly it causes substantial errors in this work. The reasoning behind this decision is justified given the overall funding dedicated to the task (two months) and the real focus of the work: can modern deterministic tools actually treat complex facilities like the ATR with heterogeneous geometry modeling. SN2ND has been demonstrated to solve problems with upwards of one trillion degrees of freedom which translates to tens of millions of finite elements, hundreds of angles, and hundreds of energy groups, resulting in a very high-fidelity model of the system unachievable by most deterministic transport codes today. A space-angle convergence study was conducted to determine the meshing and angular cubature

  18. Scoping analysis of the Advanced Test Reactor using SN2ND

    International Nuclear Information System (INIS)

    A detailed set of calculations was carried out for the Advanced Test Reactor (ATR) using the SN2ND solver of the UNIC code which is part of the SHARP multi-physics code being developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in DOE-NE. The primary motivation of this work is to assess whether high fidelity deterministic transport codes can tackle coupled dynamics simulations of the ATR. The successful use of such codes in a coupled dynamics simulation can impact what experiments are performed and what power levels are permitted during those experiments at the ATR. The advantages of the SN2ND solver over comparable neutronics tools are its superior parallel performance and demonstrated accuracy on large scale homogeneous and heterogeneous reactor geometries. However, it should be noted that virtually no effort from this project was spent constructing a proper cross section generation methodology for the ATR usable in the SN2ND solver. While attempts were made to use cross section data derived from SCALE, the minimal number of compositional cross section sets were generated to be consistent with the reference Monte Carlo input specification. The accuracy of any deterministic transport solver is impacted by such an approach and clearly it causes substantial errors in this work. The reasoning behind this decision is justified given the overall funding dedicated to the task (two months) and the real focus of the work: can modern deterministic tools actually treat complex facilities like the ATR with heterogeneous geometry modeling. SN2ND has been demonstrated to solve problems with upwards of one trillion degrees of freedom which translates to tens of millions of finite elements, hundreds of angles, and hundreds of energy groups, resulting in a very high-fidelity model of the system unachievable by most deterministic transport codes today. A space-angle convergence study was conducted to determine the meshing and angular cubature

  19. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

  20. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    Energy Technology Data Exchange (ETDEWEB)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  1. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg

    2013-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  2. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    International Nuclear Information System (INIS)

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V and V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V and V, within the next 3-4 years via the ATR Core Modeling and Simulation and V and V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  3. High flux materials testing reactor HFR Petten. Characteristics of facilities and standard irradiation devices

    International Nuclear Information System (INIS)

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to June 1978

  4. Development of the test facilities for the measurement of core flow and pressure distribution of SMART reactor

    International Nuclear Information System (INIS)

    A design of SMART reactor has been developed, of which the primary system is composed of four internal circulation pumps, a core of 57 fuel assemblies, eight cassettes of steam generators, flow mixing head assemblies, and other internal structures. Since primary design features are very different from conventional reactors, the characteristics of flow and pressure distribution are expected to be different accordingly. In order to analyze the thermal margin and hydraulic design characteristics of SMART reactor, design quantification tests for flow and pressure distribution with a preservation of flow geometry are necessary. In the present study, the design feature of the test facility in order to investigate flow and pressure distribution, named “SCOP” is described. In order to preserve the flow distribution characteristics, the SCOP is linearly reduced with a scaling ratio of 1/5. The core flow rate of each fuel assembly is measured by a venturi meter attached in the lower part of the core simulator having a similarity of pressure drop for nominally scaled flow conditions. All the 57 core simulators and 8 S/G simulators are precisely calibrated in advance of assembling in test facilities. The major parameters in tests are pressures, differential pressures, and core flow distribution. (author)

  5. Criticality safety evaluation report for 300 Area N Reactor fuel fabrication and storage facility

    Energy Technology Data Exchange (ETDEWEB)

    Schwinkendorf, K.N.

    1993-12-01

    Criticality safety support calculations for the 300 Area Fuels Manufacturing Facility have been performed to update values currently found in the criticality prevention specifications. The old safe masses and associated dimensions table has been updated with new values, which were produced using more modern computer codes that comply with Software Quality Assurance (SQA) requirements. Monte Carlo calculations, using MCNP (Carter 1991), were performed to spot check these results, obtained using the WIMS-E code (Gubbins et. al. 1982). In addition, the criticality considerations associated with certain accident, or upset, conditions were analyzed. These scenarios include fire, the bringing together of multiple safe masses into one neutronically coupled system, misstacking, and accidental interspersed moderation.

  6. Advanced Test Reactor In-Canal Ultrasonic Scanner: Experiment Design and Initial Results on Irradiated Plates

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Wachs; J. M. Wight; D. T. Clark; J. M. Williams; S. C. Taylor; D. J. Utterbeck; G. L. Hawkes; G. S. Chang; R. G. Ambrosek; N. C. Craft

    2008-09-01

    An irradiation test device has been developed to support testing of prototypic scale plate type fuels in the Advanced Test Reactor. The experiment hardware and operating conditions were optimized to provide the irradiation conditions necessary to conduct performance and qualification tests on research reactor type fuels for the RERTR program. The device was designed to allow disassembly and reassembly in the ATR spent fuel canal so that interim inspections could be performed on the fuel plates. An ultrasonic scanner was developed to perform dimensional and transmission inspections during these interim investigations. Example results from the AFIP-2 experiment are presented.

  7. The Jules Horowitz Reactor : A new high Performances European MTR (Material Testing Reactor) with modern experimental capacities : Toward an International User Facility

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is a new Material Testing Reactor (MTR) currently under construction at CEA Cadarache research centre in the south of France. It will be a major Research facility in support to the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimise and demonstrate safe operations of existing power reactors as well as to support future reactor design. It will represent also an important Research Infrastructure for scientific studies dealing with material and fuel behaviour under irradiation. The JHR will contribute also to secure the production of radioisotope for medical application. This is a key public health stake. The construction of JHR which was started in 2007 is on-going. The first operation is planned before the end of this decade.The design of the reactor will provide an essential facility supporting the programs for the nuclear energy for the next 50 years. JHR is designed to provide high neutron flux (enhancing the maximum available today in MTRs), to run highly instrumented experiments to support advanced modelling giving prediction beyond experimental points, and to operate experimental devices giving environment conditions (pressure, temperature, flux, coolant chemistry, ···) relevant for water reactors, for gas cooled thermal or fast reactors, for sodium fast reactors, ···So, the reactor will perform R and D programs for the optimization of the present generation of NPP, support the development of the next generation of NPP (mainly LWR) and also offer irradiation possibilities for future reactors. In parallel to the construction of the reactor, the preparation of an international community around JHR is continuing; this is an important topic as building and gathering a strong international community in support to MTR experiments is a key-issue for the R and D in nuclear energy field. Consequently, CEA is

  8. The Jules Horowitz Reactor : A new high Performances European MTR (Material Testing Reactor) with modern experimental capacities : Toward an International User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bignan, G.; Estrade, J. [French Atomic Energy Commission, Paris (France)

    2013-07-01

    The Jules Horowitz Reactor (JHR) is a new Material Testing Reactor (MTR) currently under construction at CEA Cadarache research centre in the south of France. It will be a major Research facility in support to the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimise and demonstrate safe operations of existing power reactors as well as to support future reactor design. It will represent also an important Research Infrastructure for scientific studies dealing with material and fuel behaviour under irradiation. The JHR will contribute also to secure the production of radioisotope for medical application. This is a key public health stake. The construction of JHR which was started in 2007 is on-going. The first operation is planned before the end of this decade.The design of the reactor will provide an essential facility supporting the programs for the nuclear energy for the next 50 years. JHR is designed to provide high neutron flux (enhancing the maximum available today in MTRs), to run highly instrumented experiments to support advanced modelling giving prediction beyond experimental points, and to operate experimental devices giving environment conditions (pressure, temperature, flux, coolant chemistry, ···) relevant for water reactors, for gas cooled thermal or fast reactors, for sodium fast reactors, ···So, the reactor will perform R and D programs for the optimization of the present generation of NPP, support the development of the next generation of NPP (mainly LWR) and also offer irradiation possibilities for future reactors. In parallel to the construction of the reactor, the preparation of an international community around JHR is continuing; this is an important topic as building and gathering a strong international community in support to MTR experiments is a key-issue for the R and D in nuclear energy field. Consequently, CEA is

  9. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    Energy Technology Data Exchange (ETDEWEB)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

  10. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    International Nuclear Information System (INIS)

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs

  11. An experimental test facility to support development of the fluoride-salt-cooled high-temperature reactor

    International Nuclear Information System (INIS)

    Highlights: • • A forced convection test loop using FLiNaK salt was constructed to support development of the FHR. • The loop is built of alloy 600, and operating conditions are prototypic of expected FHR operation. • The initial test article is designed to study pebble bed heat transfer cooled by FLiNaK salt. • The test facility includes silicon carbide test components as salt boundaries. • Salt testing with silicon carbide and alloy 600 confirmed acceptable loop component lifetime. - Abstract: The need for high-temperature (greater than 600 °C) energy transport systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The fluoride-salt-cooled high-temperature reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the fluoride-salt-cooled high-temperature reactor concept. The facility is capable of operating at up to 700 °C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system, a trace heating system, and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride-salt heat transfer inside a heated pebble bed

  12. 2015 Groundwater Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Ponds

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

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

  13. Development of sodium facilities for NSRR fast breeder reactor fuel tests. 2. Sodium capsule

    International Nuclear Information System (INIS)

    In order to commercialize fast reactors, which are expected to be long-term transmutes of plutonium and long half life radioactive wastes (such as americium) from light water reactors, safety research under accident conditions and establishment of the safety guidelines are essential. Sodium facilities, such as, (1) Purification/charging loop and test loop, and (2) Proto-type Sodium capsule, were developed and fabricated in order to pulse irradiate fast breeder reactor fuels in the Nuclear Safety Research Reactor (NSRR) of JAERI for investigation on fuel behavior under transient over-power conditions. This report presents the purpose, outlines, specifications, capabilities and operation results of the proto-type sodium capsule. Two kinds of capsule, i.e., the stagnant sodium capsule and the sodium loop, were designed to pulse irradiate Fast Reactor (FR) fuels in the NSRR under sodium cooling conditions with and without flow, respectively. Because the capsules have to safely contain chemically active sodium at high temperature and stand the pressure pulses by the sodium hummer which might be generated at fuel failure, the development of the capsule is essential for realizing the research. Thus, proto-type sodium loop, which consisted of doubly sealed container, sodium pump and flow meter, was developed. In addition, two type of flange structure for the stagnant capsule and loop was leak tested at high pressures, in order to confirm its sealing capability at room and high temperature conditions. (author)

  14. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    International Nuclear Information System (INIS)

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  15. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  16. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-01-12

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FETF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This revision reflects the 19 Oct 1999 baseline.

  17. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    BOWEN, W.W.

    1999-11-08

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FFTF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This document reflects the 1 Oct 1999 baseline.

  18. Preliminary reactor physics calculations for Phase II severe fuel damage tests in the Power Burst Facility

    International Nuclear Information System (INIS)

    This report documents the results of some preliminary scoping calculations completed in support of various recent proposals for conducting advanced severe fuel damage experiments in the Power Burst Facility (PBF). For these experiments the PBF core would be modified by installation of a large (44 cm O.D.) in-pile tube. This would allow testing of large bundles of commercial fuel rods in order to study various fuel damage mechanisms. It would also allow testing of large, homogeneous cylindrical capsules of fuel-cladding materials in order to study interactions between melted fuel and selected containment materials such as steel and concrete. Calculations of estimated attainable heat generation rates have been completed for 60 cases involving parametric changes on several conceptual capsule designs and conceptual bundle designs. These calculations were based on one-dimensonal (radial) S/sub n/ transport theory models of the PBF core and enlarged experiment space. The S/sub n/ calculations were benchmarked against corresponding three-dimensional Monte Carlo calculations for two selected cases. The results obtained in this work indicate that axial and radial average heating rates of 0.06 to 1.25 watts/cm3 per MW of PBF core power are attainable in homogeneous cylindrical test capsules of various designs. For bundles of commercial fuel rods, axial average heating rates of approximately 0.16 kW/m (0.05 kW/ft) per rod per MW of core power are attainable in 289-rod assemblies of 2.5%-enriched PWR 17 x 17-type rods. This can be increased somewhat by higher test fuel enrichments, however, criticality safety considerations are expected to limit the enrichment of such large bundles to about 7% or less

  19. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    Energy Technology Data Exchange (ETDEWEB)

    Bohachek, Randolph Charles [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-09-01

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

  20. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    International Nuclear Information System (INIS)

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

  1. The Budapest research reactor as an advanced research facility for the early 21st century

    International Nuclear Information System (INIS)

    The Budapest Research Reactor, Hungary's first nuclear facility was originally put into operation in 1959. The reactor serves for: basic and applied research, technological and commercial applications, education and training. The main goal of the reactor is to serve neutron research. This unique research possibility is used by a broad user community of Europe. Eight instruments for neutron scattering, radiography and activation analyses are already used, others (e.g. time of flight spectrometer, neutron reflectometer) are being installed. The majority of these instruments will get a much improved utilization when the cold neutron source is put into operation. In 1999 the Budapest Research Reactor was operated for 3129 full power hours in 14 periods. The normal operation period took 234 hours (starting Monday noon and finishing Thursday morning). The entire production for the year 1999 was 1302 MW days. This is a slightly reduced value, due to the installation of the cold neutron source. For the year 2000 a somewhat longer operation is foreseen (near to 4000 hours), as the cold neutron source will be operational. The operation of the reactor is foreseen at least up to the end of the first decade of the 21st century. (author)

  2. Mirror Fusion Test Facility: an intermediate device to a mirror fusion reactor

    International Nuclear Information System (INIS)

    The Mirror Fusion Test Facility (MFTF-B) now under construction at Lawrence Livermore National Laboratory represents more than an order-of-magnitude step from earlier magnetic-mirror experiments toward a future mirror fusion reactor. In fact, when the device begins operating in 1986, the Lawson criteria of ntau = 1014 cm-3.s will almost be achieved for D-T equivalent operation, thus signifying scientific breakeven. Major steps have been taken to develop MFTF-B technologies for tandem mirrors. Steady-state, high-field, superconducting magnets at reactor-revelant scales are used in the machine. The 30-s beam pulses, ECRH, and ICRH will also introduce steady-state technologies in those systems

  3. Modeling ampersand analysis of criticality-induced severe accidents during refueling for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    This paper describes work done at the Oak Ridge National Laboratory (ORNL) for evaluating the potential and resulting consequences of a hypothetical criticality accident during refueling of the 330-MW Advanced Neutron Source (ANS) research reactor. The development of an analytical capability is described. Modeling and problem formulation were conducted using concepts of reactor neutronic theory for determining power level escalation, coupled with ORIGEN and MELCOR code simulations for radionuclide buildup and containment transport Gaussian plume transport modeling was done for determining off-site radiological consequences. Nuances associated with modeling this blast-type scenario are described. Analysis results for ANS containment response under a variety of postulated scenarios and containment failure modes are presented. It is demonstrated that individuals at the reactor site boundary will not receive doses beyond regulatory limits for any of the containment configurations studied

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

    Energy Technology Data Exchange (ETDEWEB)

    K. L. Davis; D. L. Knudson; J. L. Rempe; J. C. Crepeau; S. Solstad

    2015-07-01

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

  5. Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    On October 1, 1977 work began at LLL on the Mirror Fusion Test Facility (MFTF), an advanced experimental fusion device. Scheduled for operation in late 1981, MFTF is designed as an intermediate step between present mirror machines, such as 2XIIB, and an experimental fusion reactor. This design incorporates improved technology and a better theoretical understanding of how neutral beam injection, plasma guns, and gas injection into the plasma region compensate for cooling and particle losses. With the new facility, we expect to achieve a confinement factor (n tau) of 1012 particles . sm/cm3--a tenfold increase over 2XIIB n tau values--and to increase plasma temperature to over 500 million K. The following article describes this new facility and reports on progress in some of the R and D projects that are providing the technological base for its construction

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

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, Forest Howard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-11-01

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

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

    International Nuclear Information System (INIS)

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

  8. Critical Current Test Facilities for LHC Superconducting NbTi Cable Strands

    CERN Document Server

    Boutboul, T; Denarié, C H; Oberli, L R; Richter, D

    2001-01-01

    The Rutherford-type superconducting Cu/NbTi cables of the LHC accelerator are currently mass-produced by a few industrial firms. As a part of the acceptance tests, the critical current of superconducting multifilamentary wires is systematically measured on virgin strands to qualify the wires and on extracted strands to qualify the cables. For this purpose, four test stations are in operation at CERN to measure the critical current of strands at both 4.2 K and 1.9 K in magnetic fields in the 6-11 T range. The measurement setup and procedures of these facilities are reported in this article. The quality of the critical current test is guaranteed by supervising the SPC (Statistical Process Control) charts of a reference sample. The measurement repeatability and reproducibility of the stations are found to be excellent. Moreover, the measured critical current of a strand is found to be almost independent of the test station in which the measurement is performed.

  9. Performance testing of irradiation facility rabbit system pneumatic reactor RSG-GAS using standard reference material

    International Nuclear Information System (INIS)

    The irradiation facility function test of rabbit system pneumatic (RS-5) has been done using standard reference material SRM 1633 Coal Fly ash through the sending station. Long irradiation of about 4-5 seconds. The results of qualitative analysis showed that the dominant elements listed in the certificate can be detected are Al, Ca, Mg, Si, Na, Ti, V, Mn. But only an element of Mn and Na which has a relative refractive values below 10%. And the other elements have a value relative refractive index 25% - 60%. The significant difference of value was not influenced by the position of irradiation in the reactor facility but due to the influence of the time difference between the sample and the standard count, and the half-life nuclide itself. Overall it can be said that the performance of the irradiation facility pneumatic rabbit system is good, but needs to be tested again by using different standard reference materials, in order to obtain the test results of analysis that can be trusted. (author)

  10. PROMILLE database as a part of JNC reactor physics analytical system for BFS-2 fast critical facility experiments analysis

    International Nuclear Information System (INIS)

    The PROMILLE database for experimental data from the BFS-2 fast critical facility (Institute of Physics and Power Engineering (IPPE), Russia) has been developed and embedded into the JNC reactor physics analytical system to provide a strict documentation format, a common data source for different analytical tools and a unique export interface with different reactor codes. PROMILLE should be considered not only as a database but also as a bank of interfaces because of its dynamic role in the analytical process. The database currently accepts data from the simulation materials (pellets, tubes and bars) as well as full cores descriptions. A core description involves all different unit cells forming loading elements, all types of the loading elements forming a layout and the layout itself. In fact it is a description of criticality experiments. Export interfaces for the CITATION-FBR code and the SLAROM and CASUP codes have been developed. The PROMILLE software was developed with MS Visual Basic 6.0 and the data is kept in the data tables generated with the MS Access database management system. Data for eight BFS-2 assembly configurations have been incorporated. They include BFS-58-1i1 uranium-free plutonium assembly with inert material included in its fuel matrix and also seven BFS-62 modifications simulating different stages of investigation of MOX fuel based BN-600 core. (author)

  11. Gas-cooled fast reactors. Motivation and presentation of the ENIGMA program in the MASURCA experimental critical facility

    International Nuclear Information System (INIS)

    This paper describes a new experimental physics program in support of gas cooled fast reactor (GCFR) design studies, called ENIGMA, to be performed in the MASURCA critical facility at CEA-Cadarache, France. The prospective GCFR design studies at CEA are presented, as well as the specific neutronics features needing an extension of the validation of calculation tools and nuclear data. The relevant existing experiments are briefly reviewed and the need for new experimental data is pointed out. The first phase of the proposed new experiments includes a reference core with a representative spectrum, and a series of central core substitutions involving spectrum shifts, streaming studies, low-grade Pu substitutions, innovative material (Si, Zr) substitutions. Reflector substitution zones will include elements foreseen for the reflectors (Si, Zr, C). Subsequent phases will involve larger amounts of low-grade Pu or innovative materials, and configurations representative of experimental and demonstration GCFRs. (author)

  12. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    International Nuclear Information System (INIS)

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  13. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Human Factors, Controls, and Statistics; Smith, James A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design; Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  14. Lewis Research Center's coal-fired, pressurized, fluidized-bed reactor test facility

    Science.gov (United States)

    Kobak, J. A.; Rollbuhler, R. J.

    1981-01-01

    A 200-kilowatt-thermal, pressurized, fluidized-bed (PFB) reactor, research test facility was designed, constructed, and operated as part of a NASA-funded project to assess and evaluate the effect of PFB hot-gas effluent on aircraft turbine engine materials that might have applications in stationary-power-plant turbogenerators. Some of the techniques and components developed for this PFB system are described. One of the more important items was the development of a two-in-one, gas-solids separator that removed 95+ percent of the solids in 1600 F to 1900 F gases. Another was a coal and sorbent feed and mixing system for injecting the fuel into the pressurized combustor. Also important were the controls and data-acquisition systems that enabled one person to operate the entire facility. The solid, liquid, and gas sub-systems all had problems that were solved over the 2-year operating time of the facility, which culminated in a 400-hour, hot-gas, turbine test.

  15. Technology developments for ACIGA high power test facility for advanced interferometry

    Energy Technology Data Exchange (ETDEWEB)

    Barriga, P [School of Physics, University of Western Australia, Perth, WA 6009 (Australia); Barton, M [California Institute of Technology, LIGO Project, Pasadena, CA 91125 (United States); Blair, D G [School of Physics, University of Western Australia, Perth, WA 6009 (Australia)] [and others

    2005-05-21

    The High Optical Power Test Facility for Advanced Interferometry has been built by the Australian Consortium for Interferometric Gravitational Astronomy north of Perth in Western Australia. An 80 m suspended cavity has been prepared in collaboration with LIGO, where a set of experiments to test suspension control and thermal compensation will soon take place. Future experiments will investigate radiation pressure instabilities and optical spring effects in a high power optical cavity with {approx}200 kW circulating power. The facility combines research and development undertaken by all consortium members, whose latest results are presented.

  16. Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety

  17. Implementation of advanced control system in 700 MWe fuelling machine test facility using FPGA and Industrial PCs

    International Nuclear Information System (INIS)

    Fuelling machines are required to perform ON-POWER refuelling in Pressurised Heavy Water Reactors (PHWR). Fuelling machine is an example of high level engineered and precision robotics, which performs more then thousands of complex operations with the help of mechanical actuators using electrical and fluid hydraulic powers. The complete refuelling operation is performed remotely by an automatic control system which is specifically designed to cater the step by step needs of fuelling operation along with built in safety feature to avoid any damage to man, machine and reactor safety. As the design evolution of PHWR from 220 MWe to 700 MWe has resulted in significant changes, Fuelling machines have also experienced substantial advances and design modification from each reactor. These design changes are required to be simulated and verified to ascertain the intended behaviour during actual reactor operation. In view of the above, to validate and verify the design changes of 700 MWe Fuelling machine, a fully automatic Fuelling Machine Test facility (FMTF) is developed at R and D Centre, NPCIL. This facility, with the help of three similar reactor coolant channels, is able to simulate the conditions as in actual reactor environment and performs full refuelling sequence in complete auto mode. The control system philosophy for 700 MWe FMTF was to have fully automatic and safe operating sequence with PC based architecture using minimum hardware to have compact and simplified design. This resulted in the use of Field Programmable Gate Arrays (FPGA) based hardwired logic implementation along with Industrial Computer based control and display system using Data acquisition Cards. This design has drastically reduced the complex wiring implementation of combinational and sequential logic as compared to the earlier control system of FM and resulted in a very compact and sophisticated operator console. This paper brings out the details of FMTF control system which offers numerous

  18. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  19. Feasibility of conducting a dynamic helium charging experiment for vanadium alloys in the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States); Matsui, H. [Tohoku Univ. (Japan)

    1996-10-01

    The feasibility of conducting a dynamic helium charging experiment (DHCE) for vanadium alloys in the water-cooled Advanced Test Reactor (ATR) is being investigated as part of the U.S./Monbusho collaboration. Preliminary findings suggest that such an experiment is feasible, with certain constraints. Creating a suitable irradiation position in the ATR, designing an effective thermal neutron filter, incorporating thermocouples for limited specimen temperature monitoring, and handling of tritium during various phases of the assembly and reactor operation all appear to be feasible. An issue that would require special attention, however, is tritium permeation loss through the capsule wall at the higher design temperatures (>{approx}600{degrees}C). If permeation is excessive, the reduced amount of tritium entering the test specimens would limit the helium generation rates in them. At the lower design temperatures (<{approx}425{degrees}C), sodium, instead of lithium, may have to be used as the bond material to overcome the tritium solubility limitation.

  20. ICONE-4: Proceedings. Volume 2: Advanced reactors

    International Nuclear Information System (INIS)

    The proceedings for this conference are contained in 5 volumes. This volume is divided into the following areas: advanced reactor requirements; advanced reactor design and analysis; arrangement and construction; specific reactor designs; demonstration testing; safety systems and analysis; component demonstration testing; advanced reactor containment design; licensing topics and updates; accelerator applications and spallation sources; and advanced reactor development. Separate abstracts were prepared for most papers in this volume

  1. Development of inherent core technologies for advanced reactor

    International Nuclear Information System (INIS)

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  2. Development of inherent core technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keung Koo; Noh, J.M.; Hwang, D.H. [and others

    1999-03-01

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  3. Assessment of Feasibility of the Beneficial Use of Waste Heat from the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna P. Guillen

    2012-07-01

    This report investigates the feasibility of using waste heat from the Advanced Test Reactor (ATR). A proposed glycol waste heat recovery system was assessed for technical and economic feasibility. The system under consideration would use waste heat from the ATR secondary coolant system to preheat air for space heating of TRA-670. A tertiary coolant stream would be extracted from the secondary coolant system loop and pumped to a new plate and frame heat exchanger, where heat would be transferred to a glycol loop for preheating outdoor air in the heating and ventilation system. Historical data from Advanced Test Reactor operations over the past 10 years indicates that heat from the reactor coolant was available (when needed for heating) for 43.5% of the year on average. Potential energy cost savings by using the waste heat to preheat intake air is $242K/yr. Technical, safety, and logistics considerations of the glycol waste heat recovery system are outlined. Other opportunities for using waste heat and reducing water usage at ATR are considered.

  4. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

    2012-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core

  5. Assessment of impacts at the advanced test reactor as a result of chemical releases at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    This report provides an assessment of potential impacts at the Advanced Test Reactor Facility (ATR) resulting from accidental chemical spill at the Idaho Chemical Processing Plant (ICPP). Spills postulated to occur at the Lincoln Blvd turnoff to ICPP were also evaluated. Peak and time weighted average concentrations were calculated for receptors at the ATR facility and the Test Reactor Area guard station at a height above ground level of 1.0 m. Calculated concentrations were then compared to the 15 minute averaged Threshold Limit Value - Short Term Exposure Limit (TLV-STEL) and the 30 minute averaged Immediately Dangerous to Life and Health (IDLH) limit. Several different methodologies were used to estimate source strength and dispersion. Fifteen minute time weighted averaged concentrations of hydrofluoric acid and anhydrous ammonia exceeded TLV-STEL values for the cases considered. The IDLH value for these chemicals was not exceeded. Calculated concentrations of ammonium hydroxide, hexone, nitric acid, propane, gasoline, chlorine and liquid nitrogen were all below the TLV-STEL value

  6. PSA-operations synergism for the advanced test reactor shutdown operations PSA

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) Probabilistic Safety Assessment (PSA) for shutdown operations, cask handling, and canal draining is a successful example of the importance of good PSA-operations synergism for achieving a realistic and accepted assessment of the risks and for achieving desired risk reduction and safety improvement in a best and cost-effective manner. The implementation of the agreed-upon upgrades and improvements resulted in the reductions of the estimated mean frequency for core or canal irradiated fuel uncovery events, a total reduction in risk by a factor of nearly 1000 to a very low and acceptable risk level for potentially severe events

  7. Risk-based management system development for the Advanced Test Reactor

    International Nuclear Information System (INIS)

    A Risk-Based Management System (RBMS) is being developed to facilitate the use of the Advanced Test Reactor (ATR) probabilistic risk assessment to support ATR operation. Most ATR RBMS questions can best be answered using the System Analysis and Risk Assessment System (SARA) developed at the Idaho National Engineering Laboratory. However, some applications may require employment of the other four codes used to develop and report the PRA. These four codes include the Integrated Reliability and Risk Analysis System (IRRAS), SETS, ETA-II, and the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). The ATR RBMS will evolve over three years, and will include the results of the Level 3 and external events analysis

  8. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. B. Grover

    2007-05-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing.1,2 The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment

  9. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The United States Department of Energy's Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment during

  10. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    Energy Technology Data Exchange (ETDEWEB)

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-11-30

    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  11. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    International Nuclear Information System (INIS)

    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy's resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  12. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    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) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's 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).

  13. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  14. Criticality calculation and control rods for the Westinghouse Reactor Evaluation Center facility

    International Nuclear Information System (INIS)

    This work evaluates two clean critical cores by WIMS-TRACA/CITATION codes calculation, 4 energy groups and bi dimension geometry. The first core is composed of U O2 with a clad of stainless steel and 20 absorbers Ag-In-Cd absorbers rods, the second is composed of U O2 with a clad of Zircaloy and 12 B4 C absorbers rods. (author)

  15. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Devin A. Steuhm

    2011-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore

  16. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    International Nuclear Information System (INIS)

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V and V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V and V, within the next 3-4 years via the ATR Core Modeling and Simulation and V and V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose

  17. Test and application of thermal neutron radiography facility at Xi'an pulsed reactor

    CERN Document Server

    Yang Jun; Zhao Xiang Feng; Wang Dao Hua

    2002-01-01

    A thermal neutron radiography facility at Xi'an Pulsed Reactor is described as well as its characteristics and application. The experiment results show the inherent unsharpness of BAS ND is 0.15 mm. The efficient thermal neutron n/gamma ratio is lower in not only steady state configuration but also pulsing state configuration and it is improved using Pb filter

  18. Test reactor technology

    International Nuclear Information System (INIS)

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

  19. Concepts for inventory verification in critical facilities

    International Nuclear Information System (INIS)

    Materials measurement and inventory verification concepts for safeguarding large critical facilities are presented. Inspection strategies and methods for applying international safeguards to such facilities are proposed. The conceptual approach to routine inventory verification includes frequent visits to the facility by one inspector, and the use of seals and nondestructive assay (NDA) measurements to verify the portion of the inventory maintained in vault storage. Periodic verification of the reactor inventory is accomplished by sampling and NDA measurement of in-core fuel elements combined with measurements of integral reactivity and related reactor parameters that are sensitive to the total fissile inventory. A combination of statistical sampling and NDA verification with measurements of reactor parameters is more effective than either technique used by itself. Special procedures for assessment and verification for abnormal safeguards conditions are also considered. When the inspection strategies and inventory verification methods are combined with strict containment and surveillance methods, they provide a high degree of assurance that any clandestine attempt to divert a significant quantity of fissile material from a critical facility inventory will be detected. Field testing of specific hardware systems and procedures to determine their sensitivity, reliability, and operational acceptability is recommended. 50 figures, 21 tables

  20. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO2) mixed with urania (UO2). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified

  1. Large-Scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

    International Nuclear Information System (INIS)

    The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the open-quotes flooded cavityclose quotes, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

  2. Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

    International Nuclear Information System (INIS)

    The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ''flooded cavity'', is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

  3. Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

    1994-04-01

    The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ``flooded cavity``, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications.

  4. A state-of-the-art report on the study of the nuclear reactor thermal hydraulics using integral test facilities

    International Nuclear Information System (INIS)

    Since the integral reactor (SMART) currently under development by KAERI includes distinct design features which are different from those of the conventional large scale commercial reactors, it is necessary to perform integral effect test which will be used to observe overall thermal hydraulic behavior and to verify the safety of the SMART. The integral effect test for the SMART currently promoting by the thermal hydraulic safety research team will provide experimental data to support the reactor design by the performance verification test of the reactor and safety systems, and will provide data to guarantee the safety of SMART design and to verify safety analysis codes for SMART by the integral tests. A proper scaling methodology should be applied to reflect the distinct concepts of the SMART and important physical phenomena should be preserved in this integral test facility. Thus, this report compares the conventional scaling methods and their limitation in detail, and identifies scaling distortions produced practically and examines the methods to remove or minimize the distortion. Also, by comparing in detail the design data of the conventional integral test facilities, this report summarizes comprehensively the limitations, scaling distortions and counter-measures to decrease the distortion. This report is thought to be very useful for the design and manufacturing of the integral test facility for the SMART, and is expected to be used as a guide at the conceptual design and scientific design stages of the integral test facility to simulate the operational and accidental transients to be occurred in the SMART reactor. (author). 44 refs., 27 tabs., 28 figs

  5. Mirror Fusion Test Facility magnet

    Energy Technology Data Exchange (ETDEWEB)

    Henning, C.H.; Hodges, A.J.; Van Sant, J.H.; Hinkle, R.E.; Horvath, J.A.; Hintz, R.E.; Dalder, E.; Baldi, R.; Tatro, R.

    1979-11-13

    The Mirror Fusion Test Facility (MFTF) is the largest of the mirror program experiments for magnetic fusion energy. It seeks to combine and extend the near-classical plasma confinement achieved in 2XIIB with the most advanced neutral-beam and magnet technologies. The product of ion density and confinement time will be improved more than an order of magnitude, while the superconducting magnet weight will be extrapolated from the 15 tons in Baseball II to 375 tons in MFTF. Recent reactor studies show that the MFTF will traverse much of the distance in magnet technology towards the reactor regime. Design specifics of the magnet are given. (MOW)

  6. Mirror Fusion Test Facility magnet

    International Nuclear Information System (INIS)

    The Mirror Fusion Test Facility (MFTF) is the largest of the mirror program experiments for magnetic fusion energy. It seeks to combine and extend the near-classical plasma confinement achieved in 2XIIB with the most advanced neutral-beam and magnet technologies. The product of ion density and confinement time will be improved more than an order of magnitude, while the superconducting magnet weight will be extrapolated from the 15 tons in Baseball II to 375 tons in MFTF. Recent reactor studies show that the MFTF will traverse much of the distance in magnet technology towards the reactor regime. Design specifics of the magnet are given

  7. Establishing a safety and licensing basis for generation IV advanced reactors. License by test

    International Nuclear Information System (INIS)

    The license by test approach to licensing is a novel method of licensing reactors. It provides an opportunity to deal with innovative non-water reactors in a direct way on a time scale that could permit early certification based on tests of a demonstration reactor. The uncertainties in the design and significant contributors to risk would be identified in the PRA during the design. Deterministic analysis computer codes could be tested on a real reactor. Scaling effects and associated uncertainties would be minimized. License by test is an approach that has sufficient merit to be developed and tested

  8. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

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

  9. Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2012-11-01

    This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

  10. Neural network setpoint control of an advanced test reactor experiment loop simulation

    Energy Technology Data Exchange (ETDEWEB)

    Cordes, G.A.; Bryan, S.R.; Powell, R.H.; Chick, D.R.

    1990-09-01

    This report describes the design, implementation, and application of artificial neural networks to achieve temperature and flow rate control for a simulation of a typical experiment loop in the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory (INEL). The goal of the project was to research multivariate, nonlinear control using neural networks. A loop simulation code was adapted for the project and used to create a training set and test the neural network controller for comparison with the existing loop controllers. The results for three neural network designs are documented and compared with existing loop controller action. The neural network was shown to be as accurate at loop control as the classical controllers in the operating region represented by the training set. 9 refs., 28 figs., 2 tabs.

  11. DT and DHe3 tokamak test reactor concepts using advanced, high field superconductors

    International Nuclear Information System (INIS)

    If practical high temperature superconducting ceramic magnets can be developed, there could be a significant impact on reactor design. Potential advantages include a simpler, more robust magnet design, the possibility of demountable superconducting toroidal field coils and reduced shielding requirements. The high temperature superconductors can also have very high critical fields and could provide super high field operation. This could substantially increase eta tau/sub E/ values, reduce β requirements, and improve prospects for ohmic heating to ignition. The combination of moderately high β and super high field could make DHe3 operation possible in a JET size tokamak. In this paper we discuss possibilities for test reactor designs using high temperature high field superconductors. An illustrative design has a field at the plasma of 15 T. This reduces the required β to less than 2% for DT operation. The required plasma current is 5 MA. For a reactor size of R0 = 3.4m and a = 0.6m, the neutron wall loading is 3.3 MW/m2 at β = 1.5% for DT operation and an equal amount of fusion power is produced at β = 10% for DHe3 operation. One possible mode of operation is to use ohmic heating to ignition in a DT plasma followed by thermal runaway to DHe3 temperatures. 7 refs., 1 fig., 2 tabs

  12. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  13. Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin

    2006-08-01

    Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

  14. Safety physics inter-comparison of advanced concepts of critical reactors and ADS

    International Nuclear Information System (INIS)

    Enhanced safety based on the principle of the natural ''self-defence'' is one of the most desirable features of innovative nuclear systems (critical or sub-critical) regarding both TRU transmutation and ''clean'' energy producer concepts. For the evaluation of the ''self-defence'' domain, the method of the asymptotic reactivity balance has been generalised. The promising option of Hybrids systems (that use a symbiosis of fission and spallation in sub-critical cores) which could benefit the advantages of both Accelerated Driven Systems of the traditional type and regular critical systems, has been advocated. General features of Hybrid dynamics have been presented and analysed. It was demonstrated that an external neutron source of Hybrids can expand the inherent safety potential significantly. This analysis has been applied to assess the safety physics potential of innovative concepts for prospective nuclear power both for energy producers and for transmutation. It has been found, that safety enhancement goal defines a choice of sub-criticality of Hybrids. As for energy producers with Th-fuel cycle, a significant sub-criticality level is required due to a necessity of an improvement of neutronics together with safety enhancement task. (author)

  15. Safety physics inter-comparison of advanced concepts of critical reactors and ADS

    Energy Technology Data Exchange (ETDEWEB)

    Slessarev, I. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs

    2001-07-01

    Enhanced safety based on the principle of the natural ''self-defence'' is one of the most desirable features of innovative nuclear systems (critical or sub-critical) regarding both TRU transmutation and ''clean'' energy producer concepts. For the evaluation of the ''self-defence'' domain, the method of the asymptotic reactivity balance has been generalised. The promising option of Hybrids systems (that use a symbiosis of fission and spallation in sub-critical cores) which could benefit the advantages of both Accelerated Driven Systems of the traditional type and regular critical systems, has been advocated. General features of Hybrid dynamics have been presented and analysed. It was demonstrated that an external neutron source of Hybrids can expand the inherent safety potential significantly. This analysis has been applied to assess the safety physics potential of innovative concepts for prospective nuclear power both for energy producers and for transmutation. It has been found, that safety enhancement goal defines a choice of sub-criticality of Hybrids. As for energy producers with Th-fuel cycle, a significant sub-criticality level is required due to a necessity of an improvement of neutronics together with safety enhancement task. (author)

  16. Design of Stopper of Prompt Gamma Neutron Activation Analysis Facility at China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The PGNAA facility consists of the filtered collimated neutron beam, the shielding of the whole facility, the control system, the detecting equipment and the data acquisition and analysis system. The neutron beam is filtered by a mono-crystalline bismuth filter,

  17. Technical specifications for the Oak Ridge Critical Experiments Facility

    International Nuclear Information System (INIS)

    These Technical Specifications for the Oak Ridge Critical Experiments Facility (CEF) delineate limiting conditions of operation for the facility. The CEF is used primarily for testing the High Flux Isotope Reactor (HFIR) fuel assemblies. Specifically, the Criticality Testing Unit, Liquid (CTUL), located in the CEF, is used for the HFIR fuel assembly test. The test is performed to satisfy the surveillance requirements of the HFIR Technical Specifications. The test is used to determine the water-submerged shutdown margin for each fuel assembly. 11 refs

  18. Advanced reactor licensing issues

    International Nuclear Information System (INIS)

    In July 1986 the US Nuclear Regulatory Commission issued a Policy Statement on the Regulation of Advanced Nuclear Power Plants. As part of this policy advanced reactor designers were encouraged to interact with NRC early in the design process to obtain feedback regarding licensing requirements for advanced reactors. Accordingly, the staff has been interacting with the Department of Energy (DOE) and its contractors on the review of three advanced reactor conceptual designs: one modular High Temperature Gas-Cooled Reactor (MHTGR) and two Liquid Metal Reactors (LMRs). As a result of these interactions certain safety issues associated with these advanced reactor designs have been identified as key to the licensability of the designs as proposed by DOE. The major issues in this regard are: (1) selection and treatment of accident scenarios; (2) selection of siting source term; (3) performance and reliability of reactor shutdown and decay heat removal systems; (4) need for conventional containment; (5) need for conventional emergency evacuation; (6) role of the operator; (7) treatment of balance of plant; and (8) modular approach. This paper provides a status of the NRC review effort, describes the above issues in more detail and provides the current status and approach to the development of licensing guidance on each

  19. Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data

  20. Experimental testing of reduced-scale seismic isolation bearings for the advanced liquid metal reactor

    International Nuclear Information System (INIS)

    A series of tests of reduced-scale seismic isolation bearings undertaken in support of the development of a seismic isolation concept for the Advanced Liquid Metal Reactor (ALMR) is described. A procurement specification applicable to both full-size and reduced-scale bearings was developed by the program participants and used to purchase bearings of four different designs from two manufacturers. The high-damping rubber isolators were subjected to horizontal, vertical, and failure tests designed to quantify their mechanical properties both within the range of design loads and displacements as well as to establish their margins before failure. The test results show that bearings from both manufacturers provide stable and repeatable behavior with minor variations in stiffness and damping as a function of loading frequency and load history. None of the bearings showed substantial variation in properties due to changes in axial load. All of the bearings exhibited exceptional behavior when loaded beyond the design level, with displacement margins greater than 3 and force margins greater than 4. This test program provides a thorough data-set for further analytical and experimental validations of the seismic isolation concept for the ALMR. (author)

  1. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  2. Design project of the experimental facility for testing uranium creep in the reactor

    International Nuclear Information System (INIS)

    This report contains the design for constructing the experimental device for testing metal uranium creep in the RA reactor core under defined neutron flux conditions, temperature, mechanical loads and time of irradiation. This device will be placed in one of the experimental channels in the core. This report contains physical, thermal and mechanical calculations and engineering drawings of the device

  3. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey Phillips; Charles Barnes; John Hunn

    2010-10-01

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on

  4. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    International Nuclear Information System (INIS)

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called 'AGR-1,' graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on fuel

  5. The NRU blowdown test facility commissioning program

    International Nuclear Information System (INIS)

    A major experimental program has been established at the Chalk River Nuclear Laboratories (CRL) that will provide essential data on the thermal and mechanical behaviour of nuclear fuel under abnormal reactor operating conditions and on the transient release, transport and deposition of fission product activity from severely degraded fuel. A number of severe fuel damage (SFD) experiments will be conducted within the Blowdown Test Facility (BTF) at CRL. A series of experiments are being conducted to commission this new facility prior to the SFD program. This paper describes the features and the commissioning program for the BTF. A development and testing program is described for critical components used on the reactor test section. In-reactor commissioning with a fuel assembly simulator commenced in 1989 June and preliminary results are given. The paper also outlines plans for future all-effects, in-reactor tests of CANDU-designed fuel. (author). 11 refs., 3 tabs., 7 figs

  6. Thermal hydraulic R and D of Chinese advanced reactors

    International Nuclear Information System (INIS)

    The Chinese government sponsors a program of research, development, and demonstration related to advanced reactors, both small modular reactors and larger systems. These advanced reactors encompass innovative reactor concepts, such as CAP1400 - Chinese large advanced passive pressurized water reactor, Hualong one - Chinese large advanced active and passive pressurized water reactor, ACP100 - Chinese small modular reactor, SCWR- R and D of super critical water-cooled reactor in China, CLEAR - Chinese lead-cooled fast reactor, TMSR - Chinese Thorium molten-salt reactor. The thermal hydraulic R and D of those reactors are summarised. (J.P.N.)

  7. Reactor water supplementing facility

    International Nuclear Information System (INIS)

    Condensates stored in a main condenser are introduced to a turbine-driven reactor feed water pump by way of a low pressure condensate pump, a condensate cleanup facility, a high pressure condensate pump and a low pressure feed water heater by condensate pipelines. The turbine driven feed water pump introduces feed water by way of a high pressure feed water heater to a reactor pressure vessel (RPV). Further, an auxiliary condensate pipeline having a booster pump and connected at one end to the main condenser is connected to the upstream of a motor-driven reactor feed water pump. Downstream of the turbine-driven feed water pump is connected to the downstream of the electromotive feed water pump. Then, when the condensate pump or a turbine-driven feed water pump should stop and if start of a stand-by pump is failed due to some or other reason, the motor-driven feed water pump and the booster pump are started based on a pump stop signal. With such procedures, coolants are supplied to RPV thereby enabling to ensure coolant level in the RPV. (I.N.)

  8. Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

    International Nuclear Information System (INIS)

    The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

  9. A human factors evaluation of advanced control facilities in Korea Next Generation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Seong Nam; Lee, Dong Hoon; Chung, Sung Hak; Kim, Dong Nam; Hwang, Sang Ho [Kyunghee Univ., Seoul (Korea, Republic of)

    2001-07-15

    The objectives of this study are as follows: to evaluate the impacts of advanced MMIs on operator performance; to identify new types of human errors; to present Human Factors Engineering (HFE) issues to support the safety reviews performed by the Korea Institute for Nuclear Safety. General trends in the performance measures of cognitive task demand, mental workload, and situation awareness were analyzed. The results showed that the conventional plant was superior to KNGR on the operator performance. The results of the questionnaire revealed that WDS was the most frequently used MMI resource, followed by CPS, LDP, SC, and AS. The evaluation of operator's satisfaction showed that WDS was the most satisfactory resource, followed by LDP, SC, CPS', and AS, AS was rated as the most worst resource due to inappropriate functional organization and lack of operator's visibility. Stepwise regression analyses showed that human errors of SRO and RO were mainly dominated by the cognitive behavior of 'interpretation' with WDS, while the cognitive behavior of TO was mainly dominated by 'observation' with WDS and AS. The ten HFE issues for the KNGR MCR were presented to address important design deficiencies identified in this study. The issues should be resolved to improve safety of KNGR at least up to the level of the conventional NPPs. Verification and validation activities after implementing those resolutions should be also performed to reach optimal plant safety and other operational goals.

  10. A human factors evaluation of advanced control facilities in Korea Next Generation Reactor

    International Nuclear Information System (INIS)

    The objectives of this study are as follows: to evaluate the impacts of advanced MMIs on operator performance; to identify new types of human errors; to present Human Factors Engineering (HFE) issues to support the safety reviews performed by the Korea Institute for Nuclear Safety. General trends in the performance measures of cognitive task demand, mental workload, and situation awareness were analyzed. The results showed that the conventional plant was superior to KNGR on the operator performance. The results of the questionnaire revealed that WDS was the most frequently used MMI resource, followed by CPS, LDP, SC, and AS. The evaluation of operator's satisfaction showed that WDS was the most satisfactory resource, followed by LDP, SC, CPS', and AS, AS was rated as the most worst resource due to inappropriate functional organization and lack of operator's visibility. Stepwise regression analyses showed that human errors of SRO and RO were mainly dominated by the cognitive behavior of 'interpretation' with WDS, while the cognitive behavior of TO was mainly dominated by 'observation' with WDS and AS. The ten HFE issues for the KNGR MCR were presented to address important design deficiencies identified in this study. The issues should be resolved to improve safety of KNGR at least up to the level of the conventional NPPs. Verification and validation activities after implementing those resolutions should be also performed to reach optimal plant safety and other operational goals

  11. Extreme wind induced accident sequence analysis of the advanced test reactor

    International Nuclear Information System (INIS)

    An extreme wind probabilistic risk assessment (PRA) was performed for the Department of Energy (DOE) Advanced Test Reactor (ATR) as part of the external events analysis. The ATR is located at the Idaho National Engineering Laboratory (INEL) in Idaho. The analysis included evaluation of wind fragility of several structures. As part of the analysis the impact of extreme wind on the ATR core fuel damage frequency was evaluated. Loss of commercial power was modeled as an initiating event as a function of wind velocity. Normally, the components located inside the building are not affected directly as a result of wind. However, failure of a structure can eliminate several components as a result of spatial dependency. ATR support systems are located in several structures. Two walkdowns were conducted to collect the information on structures and components and to determine the structural-components interaction. Boolean equations were developed for core fuel damage sequences which included failure of components (structures) from extreme wind, random failures and operator errors. The sequences were quantified using wind hazard curves with wind fragility of the station power, components and non-wind unavailabilities. The result showed that contribution from extreme wind was less than 4%. ATR total core damage frequency from the internal and external events is estimated to be 5.E-5/yr. The system analysis (fault trees) was performed by the EG ampersand G, Idaho Inc. and the structures and components wind fragility and sequence quantification was performed by the EQE Engineering Consultants

  12. Canister Transfer Facility Criticality Calculations

    Energy Technology Data Exchange (ETDEWEB)

    J.E. Monroe-Rammsy

    2000-10-13

    The objective of this calculation is to evaluate the criticality risk in the surface facility for design basis events (DBE) involving Department of Energy (DOE) Spent Nuclear Fuel (SNF) standardized canisters (Civilian Radioactive Waste Management System [CRWMS] Management and Operating Contractor [M&O] 2000a). Since some of the canisters will be stored in the surface facility before they are loaded in the waste package (WP), this calculation supports the demonstration of concept viability related to the Surface Facility environment. The scope of this calculation is limited to the consideration of three DOE SNF fuels, specifically Enrico Fermi SNF, Training Research Isotope General Atomic (TRIGA) SNF, and Mixed Oxide (MOX) Fast Flux Test Facility (FFTF) SNF.

  13. FFTF (FAST FLUX TEST FACILITY) REACTOR CHARACTERIZATION PROGRAM ABSOLUTE FISSION RATE MEASUREMENTS

    Energy Technology Data Exchange (ETDEWEB)

    FULLER JL; GILLIAM DM; GRUNDL JA; RAWLINS JA; DAUGHTRY JW

    1981-05-01

    Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.

  14. FFTF (Fast Flux Test Facility) Reactor Characterization Program: Absolute Fission-rate Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, J.L.; Gilliam, D.M.; Grundl, J.A.; Rawlins, J.A.; Daughtry, J.W.

    1981-05-01

    Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.

  15. State of exposure control for workers engaging in radiation works and state of radioactive waste management in nuclear reactor facilities for test and research and nuclear reactor facilities at research and development stage, fiscal year 1995

    International Nuclear Information System (INIS)

    This is the summary of the reports submitted in fiscal year 1995 by the installers of the nuclear reactor facilities for test and research or at research and development stage, conforming to the related law. The individual dose equivalent of the workers engaging in radiation works in fiscal year 1995 was sufficiently lower than the prescribed limit in all reactor facilities. As for the released quantities of gaseous and liquid wastes, the radioactive substances in the air and water outside the monitor zones never exceeded the prescribed concentration limit in all reactor facilities. In the reactor facilities, for which the target values of release control have been determined, the values were less than the targets in all cases. The increase of stored radioactive solid waste decreased as the dismantling works of the reactor auxiliary system of the nuclear powered ship 'Mutsu' were finished in fiscal year 1994. As the amount of stored radioactive solid waste approaches the installed capacity, the preservation capacity of the existing waste preservation building was increased. (K.I.)

  16. Development of CFD Approaches for Modeling Advanced Concepts of Nuclear Thermal Propulsion Test Facilities Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The project will be developing a CFD approach that can handle the additional complexities needed in a NTP testing facility when modeling the combustion processes in...

  17. Experimental facility with two-phase flow and with high concentration of non-condensable gases for research and development of emergency cooling system of advanced nuclear reactors

    International Nuclear Information System (INIS)

    The development of emergency cooling passive systems of advanced nuclear reactors requires the research of some relative processes to natural circulation, in two-phase flow conditions involving condensation processes in the presence of non-condensable gases. This work describes the main characteristics of the experimental facility called Bancada de Circulacao Natural (BCN), designed for natural circulation experiments in a system with a hot source, electric heater, a cold source, heat exchanger, operating with two-phase flow and with high concentration of noncondensable gas, air. The operational tests, the data acquisition system and the first experimental results in natural circulation are presented. The experiments are transitory in natural circulation considering power steps. The distribution of temperatures and the behavior of the flow and of the pressure are analyzed. The experimental facility, the instrumentation and the data acquisition system demonstrated to be adapted for the purposes of research of emergency cooling passive systems, operating with two-phase flow and with high concentration of noncondensable gases. (author)

  18. The high flux reactor Petten, A multi-purpose research and test facility for the future of nuclear energy

    International Nuclear Information System (INIS)

    The High Flux Reactor (HFR) at Petten, is owned by the European Commission (EC) and managed by the Institute for Advanced Materials (IAM) of the Joint Research Centre (JRC) of the EC. Its operation has been entrusted since 1962 to the Netherlands Energy Research Foundation (ECN). The HFR is one of the most powerful multi-purpose research and test reactors in the world. Together with the ECN hot cells at Petten, it has provided since three decades an integral and full complement of irradiation and examination services as required by current and future research and development for nuclear energy, industry and research organizations. Since 1963, the HFR has recognized record of consistent, reliable and high availability of more than 250 days of operation per year. The HFR has 20 in-core and 12 poolside irradiation positions, plus 12 beam tubes. With a variety of dedicated irradiation devices, and with its long-standing experience in executing small and large irradiation projects, the HFR is particularly suited for fuel, materials and components testing for all reactor lines, including thermonuclear fusion reactors. In addition, processing with neutrons and gamma rays, neutron-based research and inspection services are employed by industry and research, such as activation analysis, boron neutron capture therapy, neutron radiography and neutron diffraction. Moreover, in recent years, HFRs' mission has been broadened within the area of radioisotopes production, where, within a few years, the HFR has attained the European leadership in production volume

  19. In-situ Creep Testing Capability Development for Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

    2010-08-01

    Creep is the slow, time-dependent strain that occurs in a material under a constant strees (or load) at high temperature. High temperature is a relative term, dependent on the materials being evaluated. A typical creep curve is shown in Figure 1-1. In a creep test, a constant load is applied to a tensile specimen maintained at a constant temperature. Strain is then measured over a period of time. The slope of the curve, identified in the figure below, is the strain rate of the test during Stage II or the creep rate of the material. Primary creep, Stage I, is a period of decreasing creep rate due to work hardening of the material. Primary creep is a period of primarily transient creep. During this period, deformation takes place and the resistance to creep increases until Stage II, Secondary creep. Stage II creep is a period with a roughly constant creep rate. Stage II is referred to as steady-state creep because a balance is achieved between the work hardening and annealing (thermal softening) processes. Tertiary creep, Stage III, occurs when there is a reduction in cross sectional area due to necking or effective reduction in area due to internal void formation; that is, the creep rate increases due to necking of the specimen and the associated increase in local stress.

  20. Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs

  1. Initial high-power testing of the ATF [Advanced Toroidal Facility] ECH [electron cyclotron heating] system

    International Nuclear Information System (INIS)

    The Advanced Toroidal Facility (ATF) is a moderate aspect ratio torsatron that will utilize 53.2 GHz 200 kW Electron Cyclotron Heating (ECH) to produce nearly current-free target plasmas suitable for subsequent heating by strong neutral beam injection. The initial configuration of the ECH system from the gyrotron to ATF consists of an optical arc detector, three bellows, a waveguide mode analyzer, two TiO2 mode absorbers, two 900 miter bends, two waveguide pumpouts, an insulating break, a gate valve, and miscellaneous straight waveguide sections feeding a launcher radiating in the TE02 mode. Later, a focusing Vlasov launcher will be added to beam the ECH power to the saddle point in ATF magnetic geometry for optimum power deposition. The ECH system has several unique features; namely, the entire ECH system is evacuated, the ECH system is broadband, forward power is monitored by a newly developed waveguide mode analyzer, phase correcting miter bends will be employed, and the ECH system will be capable of operating short pulse to cw. Initial high-power tests show that the overall system efficiency is 87%. The waveguide mode analyzer shows that the gyrotron mode output consists of 13% TE01, 82.6% TE02, 2.5% TE03, and 1.9% TE04. 4 refs

  2. 2015 Annual Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Ponds

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

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

  3. The prediction of Critical Heat Flux(CHF) on the outer wall of System-Integrated Modular Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yeon, Youjin; Nam, Gyeongho; Kim, Sangnyung [Kyunghee Univ., Yongin (Korea, Republic of)

    2014-05-15

    Unlike Pressurized Water Reactor (PWR), Korea's SMART includes all components of the primary system in one pressure vessel. For this reason, if station blackout (SBO) events occur, SMART is likely to cause more severe accidents than the conventional reactor. When a nuclear meltdown, one of the serious accidents, occurs, and if the process of cooling is done by Ex-Vessel Cooling system (EVCS), this study explored as to whether EVCS is proper for SMART by calculating Critical Heat Flux (CHF) depending on the angle of reactor vessel. If the coolant is lost due to Loss-of-Coolant Accident (LOCA) and SBO in the reactor core, a core is melt due to decay heat. Then heat is transferred to the outer wall of vessel, causing Nucleate Boiling at the outer wall of vessel. A rising heat flux caused by corium could reach CHF and do damage to vessel due to Departure from Nucleate Boiling (DNB), so such incidents should be prevented. For in-vessel retention of SMART-330, this study examined the feasibility of SMART-330 by conducting a literature search of CHF correlation for the conventional commercial reactor at the outer wall of pressure vessel. The findings of the study showed that the correlation equation for SULTAN experiment was the most suitable for SMART-330. Accordingly, when the value of mass flow rate is 100kg/s and pressure is 0.2MPa, a correlation equation is simply as follows. As a result, the minimum value was about 0.5MW/m{sup 2} and the maximum was 1.1MW/m{sup 2} at the angel of 90 .deg.. Additionally, when core meltdown occurs, the maximum value of heat flux was 0.78MW/m{sup 2} at 10,000 sec after started the melting core in the reactor. (The result of heat flux was calculated heat transfer correlation inside oxide pool with number of Ra) When operating Ex-Vessel Cooling System at the outer reactor, the maximum value of CHF was about 1.1MW/m{sup 2} and it was larger than 0.78MW/m{sup 2}, the value of heat flux in the outer reactor. Therefore, this study

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

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

    International Nuclear Information System (INIS)

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high

  6. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  7. A thermal-hydraulic test rig for advanced fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    A new design of fast reactor fuel assemblies has been proposed in which the pins are supported in grids attached to the wrapper by flexible skirts. Coolant mixing is enhanced by the skirts diverting flow into the cluster of pins at each grid. There are insufficient empirical data available for the detailed design of the skirt or for the input to computer calculations of flow and heat transfer. A test rig to provide these data has been designed and built. (author)

  8. Advanced Gas Cooled Reactor Materials Program. Reducing helium impurity depletion in HTGR materials testing

    International Nuclear Information System (INIS)

    Moisture depletion in HTGR materials testing rigs has been empirically studied in the GE High Temperature Reactor Materials Testing Laboratory (HTRMTL). Tests have shown that increased helium flow rates and reduction in reactive (oxidizable) surface area are effective means of reducing depletion. Further, a portion of the depletion has been shown to be due to the presence of free C released by the dissociation of CH4. This depletion component can be reduced by reducing the helium residence time (increasing the helium flow rate) or by reducing the CH4 concentration in the test gas. Equipment modifications to reduce depletion have been developed, tested, and in most cases implemented in the HTRMTL to date. These include increasing the Helium Loop No. 1 pumping capacity, conversion of metallic retorts and radiation shields to alumina, isolation of thermocouple probes from the test gas by alumina thermowells, and substitution of non-reactive Mo-TZM for reactive metallic structural components

  9. Experimental facilities for Generation IV reactors research

    International Nuclear Information System (INIS)

    Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

  10. The second eddy current testing of zircaloy tube samples from the OECD Halden reactor project at Reactor Fuel Examination Facility, Tokai, JAERI

    International Nuclear Information System (INIS)

    The Reactor Fuel Examination Facility in Tokai/JAERI (Japan Atomic Energy Research Institute) joined to the second round robin programme on eddy current test of the Halden/IFE. In the programme, two zircaloy tube samples with some artificial defects were provided for measurements. To clarify the locations in axial and azimuthal directions, types and dimensions of the provided artificial defects, measured signals from eddy current test were analysed in comparison with the known defects on the calibration tube. As a result, fourteen defects were determined from the measurements. Then, the location, the type and the relative dimension of them were also revealed. The results of those eddy current test are described in this paper. (author)

  11. Fast reactor experiments with thorium at the PROTEUS facility

    International Nuclear Information System (INIS)

    The largescale utilization of thorium is usually linked to its introduction in fast breeder reactors and/or advanced converters. The present experiments were carried out in the zero-energy reactor facility, PROTEUS, at EIR. Six different configurations for the central fast test zone were considered in the current programme, the principal fuel/blanket materials used being in the form of rods of 15% PuO2/UO2, depleted UO2, ThO2 and Th-metal. For each configuration, measurements of the principal reaction rate ratios at the centre, as well as of reaction rate distributions across the test zone, were made. (Auth.)

  12. The Integral Test Facility Karlstein

    OpenAIRE

    Stephan Leyer; Michael Wich

    2012-01-01

    The Integral Test Facility Karlstein (INKA) test facility was designed and erected to test the performance of the passive safety systems of KERENA, the new AREVA Boiling Water Reactor design. The experimental program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests, including also the Passive Pressure Pulse Transmitter, will be performed to simulate transients and Loss of Coolant A...

  13. Flow-induced vibration test of an advanced water reactor model. Part 1: Turbulence-induced forcing function

    International Nuclear Information System (INIS)

    A 1/9 scale model of a proposed advanced water reactor was tested for flow-induced vibration. The main objectives of this test were to (1) derive an empirical equation for the turbulence forcing function which can be applied to the full-sized prototype; (2) study the effect of viscosity on the turbulence; (3) verify the superposition assumption widely used in dynamic analysis of weakly coupled fluid-shell systems; and (4) measure the shell responses to verify methods and computer programs used in the flow-induced vibration analysis of the prototype. This paper describes objectives (1), (2), and (3). Objective (4) will be discussed in a companion paper. The turbulence-induced fluctuating pressure was measured at 49 locations over the surface of a thick-walled, non-responsive scale model of the reactor vessel/core support cylinders. An empirical equation relating the fluctuating pressure, the frequency, and the distance from the inlet nozzle center line was derived to fit the test data. This equation involves only non-dimensional, fluid mechanical parameters that are postulated to represent the full-sized, geometrically similar prototype. While this postulate cannot be verified until similar measurements are taken on the full-sized unit, a similar approach using a 1/6 scale model of a commercial pressurized water reactor was verified in the mid-seventies by field measurements on the full-sized reactor

  14. Study of In-Pile test facility for fast reactor safety research: performance requirements and design features

    International Nuclear Information System (INIS)

    This paper describes a program and the main design features of a new in-pile safety facility SERAPH planned for future fast reactor safety research. The current status of R and D on technical developments in relation to the research objectives and performance requirements to the facility design is given

  15. Advanced gas cooled nuclear reactor materials evaluation and development program

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  16. Advanced gas cooled nuclear reactor materials evaluation and development program

    International Nuclear Information System (INIS)

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  18. LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

    International Nuclear Information System (INIS)

    LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

  19. Facility modernization Annular Core Research Reactor

    International Nuclear Information System (INIS)

    The Annular Core Research Reactor (ACRR) has undergone numerous modifications since its conception in response to program needs. The original reactor fuel, which was special U-ZrH TRIGA fuel designed primarily for pulsing, has been replaced with a higher pulsing capacity BeO fuel. Other advanced operating modes which use this increased capability, in addition to the pulse and steady state, have been incorporated to tailor power histories and fluences to the experiments. Various experimental facilities have been developed that range from a radiography facility to a 50 cm diameter External Fuel Ring Cavity (FREC) using 180 of the original ZrH fuel elements. Currently a digital reactor console is being produced with GA, which will give enhanced monitoring capabilities of the reactor parameters while leaving the safety-related shutdown functions with analog technology. (author)

  20. Sextant: an expert system for transient analysis of nuclear reactors and integral test facilities

    International Nuclear Information System (INIS)

    Expert systems provide a new way of dealing with the computer-aided management of nuclear plants by combining several knowledge bases and reasoning modes together with a set of numerical models for real-time analysis of transients. New development tools are required together with metaknowledge bases handling temporal hypothetical reasoning and planning. They have to be efficient and robust because during a transient, neither measurements nor models, nor scenarios are hold as absolute references. SEXTANT is a general purpose physical analyzer intended to provide a pattern and avoid duplication of general tools and knowledge bases for similar applications. It combines several knowledge bases concerning measurements, models and qualitative behavior of PWR with a mechanism of conjecture-refutation and a set of simplified models matching the current physical state. A prototype is under assessment by dealing with integral test facility transients. For its development, SEXTANT requires a powerful shell. SPIRAL is such a toolkit, oriented towards online analysis of complex processes and already used in several applications

  1. Fast reactor fuel cycle facility

    International Nuclear Information System (INIS)

    An integrated fuel cycle facility named Fast Reactor Fuel Cycle Facility (FRFCF) is planned to be set up at Kalpakkam to close the fuel cycle of the Prototype Fast Breeder Reactor (PFBR) that is already under construction there. FRFCF is the first project of its kind in India. Closure of fuel cycle of PFBR will be a significant milestone of the second stage of nuclear power programme of the Department of Atomic Energy. The facility would be ready for operation in 2014. Design work and safety review of FRFCF are presently in progress. (author)

  2. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    International Nuclear Information System (INIS)

    The Advanced Gas Reactor-1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software

  3. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  4. Destructive examination of an Alloy 600 pressurizer relief line elbow removed from an advanced test reactor loop

    International Nuclear Information System (INIS)

    While pressurizing a materials test loop at the Advanced Test Reactor (ATR), reactor operators discovered a pinhole leak in the heat affected zone of a weld that joins an Alloy 600 elbow to the nozzle of a solid Alloy 600 pressurizer. The ATR is a fuels and materials test reactor operated for the Department of Energy by EG ampersand G Idaho, Incorporated. This loop had operated for over 20 years with a deaerated PWR-type water coolant chemistry. The 1.5 inch diameter relief line interior was typically exposed to a stagnant steam phase environment under nominal operating conditions of 620 degrees F and 1800 psi. The removed elbow, the defective weld area, and a small section of the pressurizer nozzle were destructively examined. Failure was attributed to an intergranular stress corrosion cracking (SCC) mechanism. The leak path was through the heat affected zone on the elbow side of the weld, and followed a wide weld underbead area. This underbead area apparently resulted from a burn-through that occurred when the initial weld layer was made. The through-wall crack displayed a completely intergranular mode of propagation typical of SCC. Scanning Electron Microscope examinations of other areas of the weld underbead revealed microcracks at the underbead edges. Metallographic examination of the elbow material revealed that it had a microstructure that materials testing has shown to have poor resistance to SCC; in addition, grain size banding was observed

  5. CALIBRATION AND HOT TESTING OF THE ADVANCED NUCLEAR MEASUREMENT SYSTEMS USED FOR WASTE CHARACTERIZATION IN COGEMA'S NEW ACC COMPACTION FACILITY

    International Nuclear Information System (INIS)

    Spent nuclear fuel from commercial power reactors is reprocessed at the COGEMA plant in La Hague. After shearing and dissolution of the fuel assemblies, the hulls and nozzles are sent to COGEMA's new compaction facility (ACC) to reduce the final volume of waste. Technological waste generated in the reprocessing plant is also sent to the ACC facility. Compacted waste is characterized by two measurement stations: a gamma spectrometry station and an active and passive neutron measurement station. The main purpose of these measurement stations is to determine the guaranteed nuclear parameters of the compacted waste and their associated uncertainties: (1) total U and Pu masses, (2) Pu, Cm, and total alpha activities, (3) 137Cs, 90Sr-90Y,241Pu beta activities, (4) decay heat. After giving a description of the measurement stations, this paper will describe the qualification tests performed in the context of the ACC project. The extensive calibration tests performed on site with different sources and different waste matrices will be described (approximately 500 neutron and gamma experiments). Hot tests that were conducted after hot start-up at the end of 2001 and prior to the start of commercial operation will be also presented. A number of drums produced by the upstream facilities were introduced one by one into the ACC facility in order to avoid mixing of different fuel assemblies. This procedure allows comparison between characterization performed in the upstream facilities on the basis of fuel data available before processing and the measurements performed on the new ACC stations. These comparisons showed good agreement between the different methods of characterization and thus validated the innovative technologies and methods used by COGEMA for compacted waste generated by the ACC facility

  6. Critical Facilities for Coastal Geographies

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The critical facilities data are derived from the USGS Structures Inventory Database (June, 2015). The structures in the derived dataset displays aggregated totals...

  7. Operational experience on sodium deposits in KNK reactor and RSB test facility

    International Nuclear Information System (INIS)

    A specific problem of sodium-cooled reactor plants is the formation of sodium aerosols which deposit at cold sections of the plant. Formation and behaviour of sodium aerosols depend on various factors. These may show extreme different effects under conditions which first seem to be identical. Thus, it is very difficult to set up general valid rules on sodium aerosols. By operational experience gained in different plants under divers operating conditions, knowledge is drawn which corresponds well with theoretical considerations. (author)

  8. Advanced Superconducting Test Accelerator (ASTA)

    Data.gov (United States)

    Federal Laboratory Consortium — The Advanced Superconducting Test Accelerator (ASTA) facility will be based on upgrades to the existing NML pulsed SRF facility. ASTA is envisioned to contain 3 to...

  9. Screening test results on potential alternate alloys for VHTGR applications. Advanced Gas Cooled Reactor Materials Program

    International Nuclear Information System (INIS)

    General Electric is working to define and develop the materials technology which will be required for advanced very High Temperature Gas Reactors operating at primary coolant temperatures up to 9500C. The most promising application which has been identified is providing process heat for the reforming of methane. Earlier work had identified Inconel 617 and Alloy 800H as the best of the commercially available alloys for the reformer components. Since these alloys were identified, additional alloys have been developed which may offer improved performance over the above reference reformer alloys. This report presents the results obtained to date on four possible alternate alloys, Nimonic 86, Sanicro 32X, SSS-113-MA, and X 8 NiCrMoNb 16 16, which are being evaluated by General Electric for thermal stability and compatibility with HTGR helium environments. The thermal stabilities of Nimonic 86, Sanicro 32X, and X 8 NiCrMoNb 16 16 have been shown to be good out to maximum exposure times and temperatures of 6000 hours and 9500C, respectively. The thermal stability, as measured by room temperature impact strength, and post exposure ductility of the Japanese developmental alloy SSS-113-MA have been shown to be poor. Measured impact strengths and ductilities below 15 ft-lbs and 10%, respectively, have been observed for this alloy. No conclusions regarding the helium compatibility of the alloys can be made at this time because of the limited data available

  10. Real time simulator for material testing reactor

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  11. Coldleg Loss of Coolant Accident (LOCA) Analysis of the Modified Reactor Thermohydraulic Test Facility Using CATHENA Computer Code

    International Nuclear Information System (INIS)

    A LOCA analysis at coldleg of the reactor thermal hydraulic test facility using CATHENA computer code has been completely conducted. The analysis was performed by modeling the reactor thermal hydraulic test loop into generic models of the CATHENA such as PUMP, VALVE, VOLUME, ACCUMULATOR, TANK, RESERVOIR, DISCHARGE, GENERALIZED TANK, and GENHTP. The primary system was simulated at power of 1 MWatt with pressure and mass flow at 15.5 MPa and 9.4 kg/s respectively. At secondary side, feedwater flowed at 15.0 kg/s with temperature of 27 oC and pressure of 0.8 MPa. The calculation showed that during steady state the inlet and outlet temperature of test section were 121 oC and 146 oC. After calculating steady state condition, the calculation was followed by transient calculation. The transient was triggered by pipe break at coldleg with diameter of the break was 2 mm. Due to this break, the pressure decreased dramatically. When the pressure reached 4.2 MPa, the accumulator started supplying water into the system. A moment later, the pump was also tripped because of the continuing pressure drop that reached 4.0 MPa. As a consequence the coolant flow was also dropped At the coolant 40 % of normal flow, the power of heated rods then shut down. The result of calculation showed that during the transient, the maximum coolant temperature was 159 oC and the maximum temperature of heated rods was 223 oC. Based on these results, it can be concluded that during the transient, the heated rods were not in danger. (author)

  12. EURAC: the JRC proposal for an European fusion reactor materials test and development facility

    International Nuclear Information System (INIS)

    For the last 7 years we examined the use of a Spallation Neutron Source (SNS) as an altenative European Option to FMIT. For an optimized spallation neutron source design we find now for the same beam power the following design parameters: - Linear Accelerator: 600 MeV, 6 m-A-proton beam on liquid lead target - irradiation parameters: 320 dpa/year in 20 cm3 or 274 dpa/year in 31.5 cm3 6 -1 sec-1 in order to simulate the Pulsed Mode of Tokamak Power Reactors. The deflected beam can be used for other experiments

  13. The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

    OpenAIRE

    Alessandro Alemberti; Fabio Berra; Davor Grgic; Graydon Yoder; Stefano Monti; Paride Meloni; Davide Papini; Fosco Bianchi; Marco Ricotti; Roberta Ferri; Cinzia Congiu; Gustavo Cattadori; Andrea Achilli; Bojan Petrovic; Andrea Maioli

    2009-01-01

    IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be bu...

  14. Flow-induced vibration test of an advanced water reactor model. Part 2: Turbulence-induced structural response

    International Nuclear Information System (INIS)

    A 1/9-scale model flow-induced vibration test of a proposed advanced water reactor (AWR) was performed. The main objectives of the test program were: (1) to derive an empirical equation for the turbulence-induced forcing function that can be applied to the full-sized prototype; (2) to study the effect of viscosity on the turbulence forcing function generation and dissipation and to verify the superposition assumption widely used in dynamic analysis of weakly coupled fluid-shell systems; (3) to measure the shell response due to turbulence-induced excitation so that the data can be used to verify methods and computer programs used in the flow-induced vibration design analysis of the prototype. This paper describes Objective (3) of the test program

  15. Gas Test Loop Facilities Alternatives Assessment Report Rev 1

    Energy Technology Data Exchange (ETDEWEB)

    William J. Skerjanc; William F. Skerjanc

    2005-07-01

    An important task in the Gas Test Loop (GTL) conceptual design was to determine the best facility to serve as host for this apparatus, which will allow fast-flux neutron testing in an existing nuclear facility. A survey was undertaken of domestic and foreign nuclear reactors and accelerator facilities to arrive at that determination. Two major research reactors in the U.S. were considered in detail, the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR), each with sufficient power to attain the required neutron fluxes. HFIR routinely operates near its design power limit of 100 MW. ATR has traditionally operated at less than half its design power limit of 250 MW. Both of these reactors should be available for at least the next 30 years. The other major U.S. research reactor, the Missouri University Research Reactor, does not have sufficient power to reach the required neutron flux nor do the smaller research reactors. Of the foreign reactors investigated, BOR-60 is perhaps the most attractive. Monju and BN 600 are power reactors for their respective electrical grids. Although the Joyo reactor is vigorously campaigning for customers, local laws regarding transport of radioactive material mean it would be very difficult to retrieve test articles from either Japanese reactor for post irradiation examination. PHENIX is scheduled to close in 2008 and is fully booked until then. FBTR is limited to domestic (Indian) users only. Data quality is often suspect in Russia. The only accelerator seriously considered was the Fuel and Material Test Station (FMTS) currently proposed for operation at Los Alamos National Laboratory. The neutron spectrum in FMTS is similar to that found in a fast reactor, but it has a pronounced high-energy tail that is atypical of fast fission reactor spectra. First irradiation in the FMTS is being contemplated for 2008. Detailed review of these facilities resulted in the recommendation that the ATR would be the best host for the GTL.

  16. Mirror Advanced Reactor Study (MARS) final report summary

    International Nuclear Information System (INIS)

    The Mirror Advanced Reactor Study (MARS) has resulted in an overview of a first-generation tandem mirror reactor. The central cell fusion plasma is self-sustained by alpha heating (ignition), while electron-cyclotron resonance heating and negative ion beams maintain the electrostatic confining potentials in the end plugs. Plug injection power is reduced by the use of high-field choke coils and thermal barriers, concepts to be tested in the Tandem Mirror Experiment-Upgrade (TMX-U) and Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory

  17. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  18. Fiscal year 1998 multi-year work plan. Advanced reactors transition program

    International Nuclear Information System (INIS)

    The mission of the Advanced Reactors Transition program is two-fold. First, the program is to maintain the Fast Flux Test Facility (FFTF) and the Fuels and Materials Examination Facility (FMEF) in Standby to support a possible future role in the tritium production strategy. Secondly, the program is to continue deactivation activities which do not conflict with the Standby directive. On-going deactivation activities include the processing of non-usable, irradiated, FFTF components for storage or disposal; deactivation of Nuclear Energy legacy test facilities; and deactivation of the Plutonium Recycle Test Reactor (PRTR) facility, 309 Building

  19. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4∼2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure

  20. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  1. Design, development and testing of 25 NB size Accumulator Isolation Passive Valve (AIPV) for Advanced Heavy Water Reactor (AHWR)

    International Nuclear Information System (INIS)

    In Advanced Heavy Water Reactor (AHWR), Emergency Core Cooling System (ECCS) is one of the engineered safety system provided to mitigate the consequences of Loss of Coolant Accident (LOCA) in the event of a break in the pressure boundary of Main Heat Transport (MHT) circuit. High Pressure Injection System of ECCS, is designed to provide coolant injection from advanced accumulators directly into the core for 15 minutes after LOCA. The injection pipe between each accumulator and ECCS header has a newly developed passive valve called Accumulator Isolation Passive Valve (AIPV). During normal reactor operation the MHT pressure will be 70 bar and accumulator pressure will be 55 bar. With rupture of large pipe, when the MHT system pressure falls down below 50 bar, the AIPV located between the accumulators and the ECC Headers, will open to provide coolant to the core. The AIPV is a self-acting type of valve requiring no external energy, i.e.neither air nor electric supply for its actuation. The AIPV serves not only as a passive isolation device but also as a flow control device. It is a non-standard, high pressure and high temperature valve and not manufactured by the valve industry worldwide. In the process of design and development of a 200 NB prototype AIPV for AHWR, a 25 NB size AIPV was designed and developed and successfully tested at Integral Test Loop (ITL). During several experiments carried out at ITL the functional capabilities of AIPV has been proved. The in-situ calibration and testing of AIPV in the plant without removing the same has also been established. This report deals with the role of AIPV in ECCS of AHWR, its design basis, tests performed at simulated conditions and test results with analysis. (author)

  2. Reactor water injection facility

    International Nuclear Information System (INIS)

    A steam turbine and an electric generator are connected by way of a speed convertor. The speed convertor is controlled so that the number of rotation of the electric generator is constant irrespective of the speed change of the steam turbine. A shaft coupler is disposed between the turbine and the electric generator or between the turbine and a water injection pump. With such a constitution, the steam turbine and the electric generator are connected by way of the speed convertor, and since the number of revolution of the electric generator is controlled to be constant, the change of the number of rotation of the turbine can be controlled irrespective of the change of the number of rotation of the electric generator. Accordingly, the flow rate of the injection water from the water injection pump to a reactor pressure vessel can be controlled freely thereby enabling to supply stable electric power. (T.M.)

  3. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

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

  4. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

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

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

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

  6. Tests of reduced-scale seismic isolation bearings for the U.S. Advanced Liquid Metal Reactor (ALMR) program

    International Nuclear Information System (INIS)

    This paper summarizes a portion of a thorough series of tests on several different designs of reduced-scale high damping rubber isolators for the U.S. Advanced Liquid Metal Reactor (ALMR) program. A formal procurement specification has been developed by the program participants for purchasing bearings of several different scale factors and designs. The reduced-scale bearings in the specification have geometric scale factors of 1/4 and 1/8 so that dynamic tests can be performed at realistic rates, The 1/8-scale bearings also have a range of rubber layer thicknesses so that the effects of shape factor on mechanical properties may be determined. Tests of bearings from two suppliers using one of the 1/8-scale bearing designs are summarized here. The test program includes horizontal shear tests to moderate and high shear strains at a range of axial loads and frequencies, as well as vertical tests and failure tests to quantify the margin of safety in the actual design. Load-History effects including short-term stiffness reduction and long-term stiffness recovery are also under study. The primary focus of these tests is on characterizing the behavior of the compounds proposed by the bearing suppliers

  7. Tests of reduced-scale seismic isolation bearings for the U.S. Advanced Liquid Metal Reactor (ALMR) program

    Energy Technology Data Exchange (ETDEWEB)

    Clark, P.W.; Aiken, I.D.; Kelly, J.M. [Univ. of California, Berkeley, CA (United States). Earthquake Engineering Research Center; Gluekler, E.L. [General Electric Co., San Jose, CA (United States); Tajirian, F.F. [Bechtel National Corp., San Francisco, CA (United States)

    1995-12-01

    This paper summarizes a portion of a thorough series of tests on several different designs of reduced-scale high damping rubber isolators for the U.S. Advanced Liquid Metal Reactor (ALMR) program. A formal procurement specification has been developed by the program participants for purchasing bearings of several different scale factors and designs. The reduced-scale bearings in the specification have geometric scale factors of 1/4 and 1/8 so that dynamic tests can be performed at realistic rates, The 1/8-scale bearings also have a range of rubber layer thicknesses so that the effects of shape factor on mechanical properties may be determined. Tests of bearings from two suppliers using one of the 1/8-scale bearing designs are summarized here. The test program includes horizontal shear tests to moderate and high shear strains at a range of axial loads and frequencies, as well as vertical tests and failure tests to quantify the margin of safety in the actual design. Load-History effects including short-term stiffness reduction and long-term stiffness recovery are also under study. The primary focus of these tests is on characterizing the behavior of the compounds proposed by the bearing suppliers.

  8. Effect of high-density fuel loading on criticality of low enriched uranium fueled material test research reactors

    International Nuclear Information System (INIS)

    The effect of high-density fuel loading on the criticality of low enriched uranium fueled material test reactors was studied using the standard reactor physics simulation codes WIMS-D/4 and CITATION. Three strategies were considered to increase the fuel loading per plate: (1) by substituting the high-density fuel in place of low-density fuel keeping meat thickness and water channel width constant, (2) by substituting the high-density fuel in place of low-density fuel keeping fuel meat thickness fixed and optimizing the water channel width between the fuel plates and (3) by increasing the fuel meat thickness of fixed density fuel and optimizing the water channel width between the fuel plates. The fuel requirements for critical and first high power cores were determined in each case for higher fuel loadings per plate. It has been found that in the first case, core volume reduces with increasing fuel loadings per plate but requirement of fuel also increases. In the second and third case, core volume as well as fuel requirement decreases with increasing fuel loadings per plate. However in the second case, core volume reduces more rapidly than in case 3 with increasing fuel loadings per plate. Employing standard computer code PARET, steady state thermal hydraulic analysis of all these cores was performed. The thermal hydraulic analysis reveals that cores with higher densities and fixed water channel width are better from thermal hydraulic point of view and have fuel and clad temperatures within the acceptable limits. But the core with higher densities and optimum water channel width is a better choice in terms of core compaction, less 235U loading and higher neutron fluxes. Finally, the core was compacted in three steps to exploit the benefits of both types of cores. The strategy resulted in 36% reduction in the core volume, 50% increase in thermal neutron flux for irradiation and isotope production and a slight reduction in 235U loading. All this was achieved with

  9. Development of an Integrated Test Facility (ITF) for the advanced man machine interface evaluation

    International Nuclear Information System (INIS)

    An Integrated Test Facilityu (ITF) is a human factors experimental environment to evaluate an advanced Man Machine Interface(MMI) design. The ITF includes a Human Machine Simulator (HMS) comprised of a nuclear power plant function simulator, man-machine interface, experiment control station for the experiment control and design, human behavioural data measurement system, and Data Analysis and Experiment Evaluation Supporting System(DAEXESS). The most important features of ITF is to secure the flexibility and expandibility of Man Machine Interface(MMI) design to change easily the environment of experiments to accomplish the experiment's objects. In this paper, we describe a development scope and characteristics of the ITF such as, hardware and software development scope and characteristics, system thermohydraulic modelling characteristics, and experiment station characteristics for the experiment variables design and control, to be used as an experiment enviroment for the evaluation of VDU-based control room

  10. DNB measurements in the Westinghouse Critical Heat Flux Test Facility – ODEN to provide an improved correlation to increase DNB margin for the Westinghouse WWER-1000 fuel design

    International Nuclear Information System (INIS)

    Westinghouse has designed and built ODEN, a Critical Heat Flux (CHF) test loop for PWR applications. The ODEN test facility is a replacement to (and improvement upon) the well known former Heat Transfer Research Facility (HTRF) of Columbia University in New York City. The ODEN loop shares the laboratory infrastructure (power supply, heat sink and control room) with the well-known Westinghouse FRIGG BWR test loop. The ODEN loop is designed to cover Departure from Nucleate Boiling (DNB) testing needs for all types of PWR lattices in 5x5 or 6x6 rectangular geometry or in hexagonal test sections. The loop installation was completed in 2006 and qualification / benchmark testing versus HTRF data was completed in 2010. The ODEN CHF test loop was used to perform DNB measurements to provide an improved correlation to increase DNB margin for the Westinghouse WWER-1000 fuel design. The fuel was tested in an extensive thermal-hydraulic verification program with axially uniform test (typical cell) and two axially cosine tests (typical and thimble cells). The DNB measurements were performed at low to high pressures, low to high mass flows and include high steam quality conditions. The tests were carried out in a bundle configuration of 19 rods in a hexagonal array. The ODEN loop demonstrated high DNB data quality, and excellent agreement of the repeatability of the DNB data. The ODEN loop has now completed more than 1000 CHF data points. Following the DNB measurements, Westinghouse developed two DNB correlations. The WVHI correlation for predicting DNB for high flow conditions with all four loops in service operation, and the WVLO correlation for predicting DNB for low flow conditions for N-X loop operations. The WVHI and WVHLO correlations have been implemented in the Westinghouse 3-D thermalhydraulic sub-channel code VIPRE-W and used for comparative Steady-State Operation DNBR analyses as well as for comparative analyses of the Complete Loss of Flow, Under Frequency (CLOF UF

  11. In-Pile Experiment of a New Hafnium Aluminide Composite Material to Enable Fast Neutron Testing in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Douglas L. Porter; James R. Parry; Heng Ban

    2010-06-01

    A new hafnium aluminide composite material is being developed as a key component in a Boosted Fast Flux Loop (BFFL) system designed to provide fast neutron flux test capability in the Advanced Test Reactor. An absorber block comprised of hafnium aluminide (Al3Hf) particles (~23% by volume) dispersed in an aluminum matrix can absorb thermal neutrons and transfer heat from the experiment to pressurized water cooling channels. However, the thermophysical properties, such as thermal conductivity, of this material and the effect of irradiation are not known. This paper describes the design of an in-pile experiment to obtain such data to enable design and optimization of the BFFL neutron filter.

  12. Test facilities and on-line instrumentation capabilities for core component materials investigations at the HALDEN reactor project

    International Nuclear Information System (INIS)

    This paper describes experimental facilities at the Halden Reactor Project, dedicated to studying the behaviour of LWR core component materials in environments simulating those of operating nuclear power plants in terms of thermal hydraulic, neutronic and water chemistry conditions. The majority of the materials investigations make use of in-pile measurements. On-line monitoring techniques, such as the reversing dc potential drop method for crack propagation monitoring and the use of Linear Variable Differential Transformers (LVDTs) for crack initiation and creep and stress relaxation studies, are described and results from studies employing these instrumentation methods are presented. The development of cracks due to the mechanism of irradiation assisted stress corrosion cracking (IASCC) is a process that affects the lifetime of nuclear power plants and there is a need both for the industry and safety authorities to have reliable materials data for use in safety assessments. IASCC of in-core components is a cause for concern for both BWRs and PWRs as reactors age, with components such as the core shroud and top guide in BWRs and the baffle former bolts in PWRs having experienced intergranular cracking attributed to IASCC. The main objective the crack growth studies that have been performed at Halden for a number of years are to generate long-term crack growth rate data for irradiated materials in typical LWR conditions. The effects of fluence, radiation hardening and applied stress intensity level on cracking are also addressed. Between four to six Compact Tension (CT) specimens, equipped with pressurised bellows for load application and instrumented for crack propagation monitoring with the reversing DC potential drop (DCPD) method, are accommodated in the test assemblies. The specimens are prepared from irradiated 304, 316 and 347 SS (with fluences ranging from 7 x 1019 to 2.5 x 1022 n/cm2 (> 1 MeV)) taken from commercial reactor core components. Examples of

  13. Computational analysis of irradiation facilities at the JSI TRIGA reactor.

    Science.gov (United States)

    Snoj, Luka; Zerovnik, Gašper; Trkov, Andrej

    2012-03-01

    Characterization and optimization of irradiation facilities in a research reactor is important for optimal performance. Nowadays this is commonly done with advanced Monte Carlo neutron transport computer codes such as MCNP. However, the computational model in such calculations should be verified and validated with experiments. In the paper we describe the irradiation facilities at the JSI TRIGA reactor and demonstrate their computational characterization to support experimental campaigns by providing information on the characteristics of the irradiation facilities. PMID:22154389

  14. Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico

    International Nuclear Information System (INIS)

    The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program

  15. Laser solenoid radiation test facility

    International Nuclear Information System (INIS)

    The Laser Solenoid Radiation Test Facility (LSRTF) is a concept based on a pulsed plasma source of neutrons, alpha particles, and bremsstrahlung and is characterized by a moderate radiation flux and a large test sample volume. The LSRTF is intermediate in its size, technology, and availability (1985-1990), and consequently has potential for bridging the gap between small present day accelerator-target sources and a large pulsed plasma engineering research facility in the 1990's. It also has important potential as a compact engineering test reactor for realistic operational testing of integrated subsystems for a linear fusion reactor. Its design, performance and operating characteristics are discussed in the present paper. The necessary development programs to bring such a facility into timely operation are also described. (Auth.)

  16. The Integral Test Facility Karlstein

    Directory of Open Access Journals (Sweden)

    Stephan Leyer

    2012-01-01

    Full Text Available The Integral Test Facility Karlstein (INKA test facility was designed and erected to test the performance of the passive safety systems of KERENA, the new AREVA Boiling Water Reactor design. The experimental program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests, including also the Passive Pressure Pulse Transmitter, will be performed to simulate transients and Loss of Coolant Accident scenarios at the test facility. The INKA test facility represents the KERENA Containment with a volume scaling of 1 : 24. Component heights and levels are in full scale. The reactor pressure vessel is simulated by the accumulator vessel of the large valve test facility of Karlstein—a vessel with a design pressure of 11 MPa and a storage capacity of 125 m3. The vessel is fed by a benson boiler with a maximum power supply of 22 MW. The INKA multi compartment pressure suppression Containment meets the requirements of modern and existing BWR designs. As a result of the large power supply at the facility, INKA is capable of simulating various accident scenarios, including a full train of passive systems, starting with the initiating event—for example pipe rupture.

  17. The Integral Test Facility Karlstein

    International Nuclear Information System (INIS)

    The Integral Test Facility Karlstein (INKA) test facility was designed and erected to test the performance of the passive safety systems of KERENA, the new AREVA Boiling Water Reactor design. The experimental program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests, including also the Passive Pressure Pulse Transmitter, will be performed to simulate transients and Loss of Coolant Accident scenarios at the test facility. The INKA test facility represents the KERENA Containment with a volume scaling of 1?:?24. Component heights and levels are in full scale. The reactor pressure vessel is simulated by the accumulator vessel of the large valve test facility of Karlstein vessel with a design pressure of 11?MPa and a storage capacity of 125?m3. The vessel is fed by a benson boiler with a maximum power supply of 22?MW. The INKA multi compartment pressure suppression Containment meets the requirements of modern and existing BWR designs. As a result of the large power supply at the facility, INKA is capable of simulating various accident scenarios, including a full train of passive systems, starting with the initiating event for example pipe rupture.

  18. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The overall objective of the Phase 1 effort was to demonstrate the technical feasibility of the Advanced Carbothermal Electric (ACE) Reactor concept. Unlike...

  19. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

  20. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  1. Neutron beam facilities at the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    The exciting development for Australia is the construction of a modern state-of-the-art 20-MW Replacement Research Reactor which is currently under construction to replace the aging reactor (HIFAR) at ANSTO in 2006. To cater for advanced scientific applications, the replacement reactor will provide not only thermal neutron beams but also a modern cold-neutron source moderated by liquid deuterium at approximately -250 deg C, complete with provision for installation of a hot-neutron source at a later stage. The latest 'supermirror' guides will be used to transport the neutrons to the Reactor Hall and its adjoining Neutron Guide Hall where a suite of neutron beam instruments will be installed. These new facilities will expand and enhance ANSTO's capabilities and performance in neutron beam science compared with what is possible with the existing HIFAR facilities, and will make ANSTO/Australia competitive with the best neutron facilities in the world. Eight 'leading-edge' neutron beam instruments are planned for the Replacement Research Reactor when it goes critical in 2006, followed by more instruments by 2010 and beyond. Up to 18 neutron beam instruments can be accommodated at the Replacement Research Reactor, however, it has the capacity for further expansion, including potential for a second Neutron Guide Hall. The first batch of eight instruments has been carefully selected in conjunction with a user group representing various scientific interests in Australia. A team of scientists, engineers, drafting officers and technicians has been assembled to carry out the Neutron Beam Instrument Project to successful completion. Today, most of the planned instruments have conceptual designs and are now being engineered in detail prior to construction and procurement. A suite of ancillary equipment will also be provided to enable scientific experiments at different temperatures, pressures and magnetic fields. This paper describes the Neutron Beam Instrument Project and gives

  2. Los Alamos Critical Experiments Facility

    International Nuclear Information System (INIS)

    The Critical Experiments Facility of the Los Alamos National Laboratory has been in existence for 45 years. In that period of time, thousands of measurements have been made on assemblies containing every fissionable material in various configurations that included bare metal and compounds of the nitrate, sulfate, fluoride, carbide, and oxide. Techniques developed or applied include Rossi-α, source-jerk, rod oscillator, and replacement measurements. Many of the original measurements of delay neutrons were performed at the site, and a replica of the Hiroshima weapon was operated at steady state to assist in evaluating the relative biological effectiveness (RBE) of neutrons. Solid, liquid, and gas fissioning systems were run at critical. Operation of this original critical facility has demonstrated the margin of safety that can be obtained through remote operation. Eight accidental excursions have occurred on the site, ranging from 1.5 x 1016 to 1.2 x 1017 fissions, with no significant exposure to personnel or damage to the facility beyond the machines themselves -- and in only one case was the machine damaged beyond further use. The present status of the facility, operating procedures, and complement of machines will be described in the context of programmatic activity. New programs will focus on training, validation of criticality alarm systems, experimental safety assessment of process applications, and dosimetry. Special emphasis will be placed on the incorporation of experience from 45 years of operation into present procedures and programs. 3 refs

  3. Five years operating experience at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    The Fast Flux Test Facility (FFTF) is a 400 Mw(t), loop-type, sodium-cooled, fast neutron reactor. It is operated by the Westinghouse Hanford Company for the United States Department of Energy at Richland, Washington. The FFTF is a multipurpose test reactor used to irradiate fuels and materials for programs such as Liquid Metal Reactor (LMR) research, fusion research, space power systems, isotope production and international research. FFTF is also used for testing concepts to be used in Advanced Reactors which will be designed to maximize passive safety features and not require complex shutdown systems to assure safe shutdown and heat removal. The FFTF also provides experience in the operation and maintenance of a reactor having prototypic components and systems typical of large LMR (LMFBR) power plants. The 5 year operational performance of the FFTF reactor is discussed in this report. 6 refs., 10 figs., 2 tabs

  4. Code qualification of structural materials for AFCI advanced recycling reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

  5. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  6. Textiles Performance Testing Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — The Textiles Performance Testing Facilities has the capabilities to perform all physical wet and dry performance testing, and visual and instrumental color analysis...

  7. GPS Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Global Positioning System (GPS) Test Facility Instrumentation Suite (GPSIS) provides great flexibility in testing receivers by providing operational control of...

  8. Advanced Fission Reactor Program objectives

    International Nuclear Information System (INIS)

    The objective of an advanced fission reactor program should be to develop an economically attractive, safe, proliferation-resistant fission reactor. To achieve this objective, an aggressive and broad-based research and development program is needed. Preliminary work at Brookhaven National Laboratory shows that a reasonable goal for a research program would be a reactor combining as many as possible of the following features: (1) initial loading of uranium enriched to less than 15% uranium 235, (2) no handling of fuel for the full 30-year nominal core life, (3) inherent safety ensured by core physics, and (4) utilization of natural uranium at least 5 times as efficiently as light water reactors

  9. FENIX [Fusion ENgineering International eXperimental]: A test facility for ITER [International Thermonuclear Experimental Reactor] and other new superconducting magnets

    International Nuclear Information System (INIS)

    The Fusion ENgineering International eXperimental (FENIX) Test Facility which is nearing completion at Lawrence Livermore National Laboratory, is a 76-t set of superconducting magnets housed in a 4-m-diameter cryostat. It represents a significant step toward meeting the testing needs for the development of superconductors appropriate for large-scale magnet applications such as the International Thermonuclear Experimental Reactor (ITER). The magnet set is configured to allow radial access to the 0.4-m-diameter high-field region where maximum fields up to 14 T will be provided. The facility is fitted with a thermally isolated test well with a port to the high-field region that allows insertion and removal of test conductors without disturbing the cryogenic environment of the magnets. It is expected that the facility will be made available to magnet developers internationally, and this paper discusses its general design features, its construction, and its capabilities

  10. Structural materials challenges for advanced reactor systems

    Science.gov (United States)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

  11. Criticality of spent reactor fuel

    International Nuclear Information System (INIS)

    The storage capacity of spent reactor fuel pools can be greatly increased by consolidation. In this process, the fuel rods are removed from reactor fuel assemblies and are stored in close-packed arrays in a canister or skeleton. An earlier study examined criticality consideration for consolidation of Westinghouse fuel, assumed to be fresh, in canisters at the Millstone-2 spent-fuel pool and in the General Electric IF-300 shipping cask. The conclusions were that the fuel rods in the canister are so deficient in water that they are adequately subcritical, both in normal and in off-normal conditions. One potential accident, the water spill event, remained unresolved in the earlier study. A methodology is developed here for spent-fuel criticality and is applied to the water spill event. The methodology utilizes LEOPARD to compute few-group cross sections for the diffusion code PDQ7, which then is used to compute reactivity. These codes give results for fresh fuel that are in good agreement with KENO IV-NITAWL Monte Carlo results, which themselves are in good agreement with continuous energy Monte Carlo calculations. These methodologies are in reasonable agreement with critical measurements for undepleted fuel

  12. Rocky Flats CAAS System Recalibrated, Retested, and Analyzed to Install in the Criticality Experiments Facility at the Nevada Test Site

    OpenAIRE

    Kim, S.; Heinrichs, D; Biswas, D.; Huang, S.; G. Dulik; J. Scorby; Boussoufi, M.; B. Liu; Wilson, R.

    2009-01-01

    Neutron detectors and control panels transferred from Rocky Flats Plant (RFP) were recalibrated and retested for redeployment to the CEF. Testing and calibration were successful with no failure to any equipment. Detector sensitivity was tested at the TRIGA reactor, and the response to thermal neutron flux was satisfactory. MCNP calculated minimum fission yield (~ 2 × 1015 fissions) was applied to determine the thermal flux at selected detector positions at the CEF. Thermal flux levels were...

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne Leland [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

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

  15. ''ROSTO'' Organic-moderated Critical Facility

    International Nuclear Information System (INIS)

    As a part of the Italian Project for an Organic Moderated Nuclear Power Plant, the National Commission for Nuclear Energy (CNEN) has installed a critical facility, known as ROSPO, planned to perform criticality studies of the nuclear characteristics of cores to be tested in a 60 MW(t) prototype reactor (PRO). The first core loading consists of MTR-type fuel elements, with flat, 90% enriched, stainless-steel clad plates; the main physical question to be answered concerns the possibility of obtaining, with this core, a proper reactivity excess, in order to compensate for temperature and poisoning effects, and to ensureta convenient lifetime to the power reactor. The variable parameter is the uranium/stainless steel ratio in the fuel element, and the experiments are performed by studying cores where some of the 16 active plates of each fuel element are substituted by dummy (stainless steel) plates. For three of these cores (0,3 and 5 dummy plates per element respectively), detailed calculations have been carried out to obtain flux plots, reactivity effects and keff versus number of elements. The calculation was based upon a modified two-group theory that accounts for fissions in the epithermal zone. The following codes were used: THESIS for thennal, and MUFT 4 for epithermal spectra; FLIP 1 (transport theory) and TUT-T5 (Monte Carlo) for fluxes in the unit cell; PDQ-O2 coupled to WANDA 4 (diffusion theory in xy and slab geometries) for reactivities and macroscopic fluxes. The results are shown and compared with those obtained for the same cores by AGIP Nucleare and by the CNEN Bologna Computation Centre; in spite of the analogy of the methods used, large discrepancies were observed, which are thought to derive from differences in the libraries of cross-sections. This seems to confirm the necessity of critical experiments as the most valid help in choosing and checking calculation methods, nuclear codes and libraries. The measurement of keff versus the number of fuel

  16. 2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond

    Energy Technology Data Exchange (ETDEWEB)

    Mike Lewis

    2012-02-01

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance and other issues Discussion of the facility's environmental impacts During the 2011 permit year, approximately 166 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

  17. Process Knowledge Summary Report for Advanced Test Reactor Complex Contact-Handled Transuranic Waste Drum TRA010029

    Energy Technology Data Exchange (ETDEWEB)

    B. R. Adams; R. P. Grant; P. R. Smith; J. L. Weisgerber

    2013-09-01

    This Process Knowledge Summary Report summarizes information collected to satisfy the transportation and waste acceptance requirements for the transfer of one drum containing contact-handled transuranic (TRU) actinide standards generated by the Idaho National Laboratory at the Advanced Test Reactor (ATR) Complex to the Advanced Mixed Waste Treatment Project (AMWTP) for storage and subsequent shipment to the Waste Isolation Pilot Plant for final disposal. The drum (i.e., Integrated Waste Tracking System Bar Code Number TRA010029) is currently stored at the Materials and Fuels Complex. The information collected includes documentation that addresses the requirements for AMWTP and applicable sections of their Resource Conservation and Recovery Act permits for receipt and disposal of this TRU waste generated from ATR. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for this TRU waste originating from ATR.

  18. Advanced Safeguards Approaches for New Reprocessing Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Richard; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-06-24

    U.S. efforts to promote the international expansion of nuclear energy through the Global Nuclear Energy Partnership (GNEP) will result in a dramatic expansion of nuclear fuel cycle facilities in the United States. New demonstration facilities, such as the Advanced Fuel Cycle Facility (AFCF), the Advanced Burner Reactor (ABR), and the Consolidated Fuel Treatment Center (CFTC) will use advanced nuclear and chemical process technologies that must incorporate increased proliferation resistance to enhance nuclear safeguards. The ASA-100 Project, “Advanced Safeguards Approaches for New Nuclear Fuel Cycle Facilities,” commissioned by the NA-243 Office of NNSA, has been tasked with reviewing and developing advanced safeguards approaches for these demonstration facilities. Because one goal of GNEP is developing and sharing proliferation-resistant nuclear technology and services with partner nations, the safeguards approaches considered are consistent with international safeguards as currently implemented by the International Atomic Energy Agency (IAEA). This first report reviews possible safeguards approaches for the new fuel reprocessing processes to be deployed at the AFCF and CFTC facilities. Similar analyses addressing the ABR and transuranic (TRU) fuel fabrication lines at AFCF and CFTC will be presented in subsequent reports.

  19. Ouellette Thermal Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Thermal Test Facility is a joint Army/Navy state-of-the-art facility (8,100 ft2) that was designed to: Evaluate and characterize the effect of flame and thermal...

  20. AGING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    The purpose of this design calculation is to revise and update the previous criticality calculation for the Aging Facility (documented in BSC 2004a). This design calculation will also demonstrate and ensure that the storage and aging operations to be performed in the Aging Facility meet the criticality safety design criteria in the ''Project Design Criteria Document'' (Doraswamy 2004, Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''SNF Aging System Description Document'' (BSC [Bechtel SAIC Company] 2004f, p. 3-12). The scope of this design calculation covers the systems and processes for aging commercial spent nuclear fuel (SNF) and staging Department of Energy (DOE) SNF/High-Level Waste (HLW) prior to its placement in the final waste package (WP) (BSC 2004f, p. 1-1). Aging commercial SNF is a thermal management strategy, while staging DOE SNF/HLW will make loading of WPs more efficient (note that aging DOE SNF/HLW is not needed since these wastes are not expected to exceed the thermal limits form emplacement) (BSC 2004f, p. 1-2). The description of the changes in this revised document is as follows: (1) Include DOE SNF/HLW in addition to commercial SNF per the current ''SNF Aging System Description Document'' (BSC 2004f). (2) Update the evaluation of Category 1 and 2 event sequences for the Aging Facility as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004c, Section 7). (3) Further evaluate the design and criticality controls required for a storage/aging cask, referred to as MGR Site-specific Cask (MSC), to accommodate commercial fuel outside the content specification in the Certificate of Compliance for the existing NRC-certified storage casks. In addition, evaluate the design required for the MSC that will accommodate DOE SNF/HLW. This design calculation will achieve the objective of providing the criticality safety results to support the preliminary design of the Aging

  1. Los Alamos Critical Assemblies Facility

    International Nuclear Information System (INIS)

    The Critical Assemblies Facility of the Los Alamos National Laboratory has been in existence for thirty-five years. In that period, many thousands of measurements have been made on assemblies of 235U, 233U, and 239Pu in various configurations, including the nitrate, sulfate, fluoride, carbide, and oxide chemical compositions and the solid, liquid, and gaseous states. The present complex of eleven operating machines is described, and typical applications are presented

  2. Characteristics and facilities of a 3MW LEU fuelled TRIGA reactor

    International Nuclear Information System (INIS)

    A 3 MW TRIGA reactor fuelled with low enriched uranium having 19.7 % U-235 and 20 wt% Uranium and Zirconium Hydride, has been installed and recently made critical at a research laboratory of the Bangladesh Atomic Energy Commission. This paper describes the basic design, low and high power test results and the facilities of the reactor. The details of the core configuration of the initial criticality with 50 elements and the final core with 100 elements at 3 MW power are discussed. The available experimental facilities are also described briefly. (author)

  3. Evaluation of selected ex-reactor accidents related to the tritium and medical isotope production missions at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    The Fast Flux Test Facility (FFTF) has been proposed as a production facility for tritium and medical isotopes. A range of postulated accidents related to ex-reactor irradiated fuel and target handling were identified and evaluated using new source terms for the higher fuel enrichment and for the tritium and medical isotope targets. In addition, two in-containment sodium spill accidents were re-evaluated to estimate effects of increased fuel enrichment and the presence of the Rapid Retrieval System. Radiological and toxicological consequences of the analyzed accidents were found to be well within applicable risk guidelines

  4. Local AREA networks in advanced nuclear reactors

    International Nuclear Information System (INIS)

    The report assesses Local Area Network Communications with a view to their application in advanced nuclear reactor control and protection systems. Attention is focussed on commercially available techniques and systems for achieving the high reliability and availability required. A basis for evaluating network characteristics in terms of broadband or baseband type, medium, topology, node structure and access method is established. The reliability and availability of networks is then discussed. Several commercial networks are briefly assessed and a distinction made between general purpose networks and those suitable for process control. The communications requirements of nuclear reactor control and protection systems are compared with the facilities provided by current technology

  5. Two-phase 3 x 3 rod bundle test facility for post-critical heat flux boiling

    International Nuclear Information System (INIS)

    This report describes the rod bundle post-CHF tests in progress and the test facility at Lehigh University. The mechanical and electrical design of the experimental facility and the iterative process used to arrive at the choices made for the design are described in detail. The test facility consists of a nine (3 x 3) rod bundle in a square shroud which form the test section together with the hot patches at the top and bottom ends. The rods and the hot patches are electrically heated while the shroud is radiatively heated. The test section includes instrumentation to measure the vapor superheat temperature and pressure drop upstream and downstream of a rod gap spacer. This is the first application of the hot patch technique for generating post-CHF conditions in a rod bundle and thus quasi-steady-state tests are being thought of as a backup procedure for conducting these post-CHF heat transfer tests. The test section is part of a well instrumented recirculating loop to generate the desired post-CHF conditions. The other major components of the heat transfer loop include the surge tank, pumps, boiler, separation tank and condenser

  6. FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-10-31

    This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

  7. Mark 1 Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Mark I Test Facility is a state-of-the-art space environment simulation test chamber for full-scale space systems testing. A $1.5M dollar upgrade in fiscal year...

  8. Structural Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Provides a wide variety of testing equipment, fixtures and facilities to perform both unique aviation component testing as well as common types of materials testing...

  9. Facility for Advanced Accelerator Experimental Tests at SLAC (FACET) Conceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Amann, J.; Bane, K.; /SLAC

    2009-10-30

    This Conceptual Design Report (CDR) describes the design of FACET. It will be updated to stay current with the developing design of the facility. This CDR begins as the baseline conceptual design and will evolve into an 'as-built' manual for the completed facility. The Executive Summary, Chapter 1, gives an introduction to the FACET project and describes the salient features of its design. Chapter 2 gives an overview of FACET. It describes the general parameters of the machine and the basic approaches to implementation. The FACET project does not include the implementation of specific scientific experiments either for plasma wake-field acceleration for other applications. Nonetheless, enough work has been done to define potential experiments to assure that the facility can meet the requirements of the experimental community. Chapter 3, Scientific Case, describes the planned plasma wakefield and other experiments. Chapter 4, Technical Description of FACET, describes the parameters and design of all technical systems of FACET. FACET uses the first two thirds of the existing SLAC linac to accelerate the beam to about 20GeV, and compress it with the aid of two chicanes, located in Sector 10 and Sector 20. The Sector 20 area will include a focusing system, the generic experimental area and the beam dump. Chapter 5, Management of Scientific Program, describes the management of the scientific program at FACET. Chapter 6, Environment, Safety and Health and Quality Assurance, describes the existing programs at SLAC and their application to the FACET project. It includes a preliminary analysis of safety hazards and the planned mitigation. Chapter 7, Work Breakdown Structure, describes the structure used for developing the cost estimates, which will also be used to manage the project. The chapter defines the scope of work of each element down to level 3.

  10. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  11. FBR related test facilities data base

    International Nuclear Information System (INIS)

    The questionnaire of main specifications, test performance and features of each FBR related test facility in the O-arai Engineering Center were made from 2001 to 2002. This report equipped these questionnaires with database. Two tables list 134 facilities. These related test facilities contains the safety test, thermal hydraulics test, test facilities for structure, reactor, Na related test, irradiation rig, fuel monitoring facility and apparatus and others (failed fuel detection and location, helium accumulation fluence monitor measurement system, inductively coupled plasma mass spectrometer, laser resonance ionization mass spectrometry system, pressurized resistance welding equipment, fuel inspection system and inductively coupled plasma mass spectrometer). This report contains all questionnaires as data. (S.Y.)

  12. Directions in advanced reactor technology

    International Nuclear Information System (INIS)

    Successful nuclear power plant concepts must simultaneously performance in terms of both safety and economics. To be attractive to both electric utility companies and the public, such plants must produce economical electric energy consistent with a level of safety which is acceptable to both the public and the plant owner. Programs for reactor development worldwide can be classified according to whether the reactor concept pursues improved safety or improved economic performance as the primary objective. When improved safety is the primary goal, safety enters the solution of the design problem as a constraint which restricts the set of allowed solutions. Conversely, when improved economic performance is the primary goal, it is allowed to be pursued only to an extent which is compatible with stringent safety requirements. The three major reactor coolants under consideration for future advanced reactor use are water, helium and sodium. Reactor development programs focuses upon safety and upon economics using each coolant are being pursued worldwide. These programs are discussed

  13. Advanced Microscopy Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides a facility for high-resolution studies of complex biomolecular systems. The goal is an understanding of how to engineer biomolecules for various...

  14. Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity - Irradiation Tests of Np and Np-U Samples in Experimental Fast Reactor Joyo (JAEA) and Advanced Thermal Reactor ATR (INL)

    International Nuclear Information System (INIS)

    A project of Produce Protected Plutonium (P3) was proposed by Tokyo-Tech as a part of non-proliferation research for Plutonium (Pu) utilization to nuclear reactor. The project is to reach the production of inherently protected Pu by addition 237Np to Uranium (U) fuel. In order to validate this P3 concept, two irradiation tests were performed. Experimental determination of Pu isotopes in 237Np samples irradiated in the experimental fast reactor Joyo was done to evaluate 238Pu production from 237Np under the fast neutron spectra. The amount of 238Pu in the irradiated 237Np samples was determined by a radiochemical analysis in Alpha-gamma Facility of JAEA. The produced 238Pu in the samples was found to depend on the neutron spectrum, ranging from that of a typical fast reactor to a nearly epi-thermal spectrum. The fast reactor can potentially control the 238Pu production from 237Np by the spectrum shift in the different irradiation position. Within the framework of the P3, the fast reactor can make roles of protected Pu production which can be better performed in the reflector region, the ratio of 238Pu is achieved up to around 90% produced from 237Np. On the other hand, 2, 5 and 10% 237Np containing U samples were also irradiated in Advanced Thermal Reactor of INL to evaluate the 238Pu production under thermal neutron region. The irradiation condition and its loading position of samples were fixed based on a calculation with MCWO (MCNP Coupled with ORIGEN2) code. The fuel specimens were removed from the core at 100, 200 and 300 effective full power days (EFPD), and then post irradiation examination was completed at Chemical lab. in MFC of INL. For the samples after 300 EFPD irradiation, Np depletion were about 60 % for 2% Np-U samples, about 50% for 5 and 10 % Np-U samples. The 238Pu to Pu ratio was about 20%, 30% and 45% for 2%, 5%, and 10% Np-U samples, respectively. The neutronics calculation results were coincident with the experimental ones. Acknowledgments: The

  15. The need to address the larger universe of HEU-fueled reactors, including: Critical assemblies, pulsed reactors and propulsion reactors

    International Nuclear Information System (INIS)

    The RERTR program has focused on ending shipments of HEU fuel to research reactors. Highest priority has been given to reactors with steady thermal powers ≥ 1 megawatt. Since the cores of critical assemblies and pulsed reactors can contain huge amounts of HEU, they should be a second focus. Also, since many aging and specialized HEU-fuelled reactors may no longer be needed, more emphasis should be given to initiatives that could assist in their shutdown and decommissioning, including providing access to regional reactors with superior facilities. HEU-fuelled ship-propulsion reactors should also be addressed. Russia's civilian icebreaker reactors are of particular interest because their fuel design is considered less sensitive than that of naval reactor fuel. Moreover, Russia's KLT-40 icebreaker reactor is being adapted for a floating nuclear power plant and LEU icebreaker fuel could be used for converting Russian research reactors such as PIK and SM-3, that operate at power-reactor temperatures. (author)

  16. Advanced Burner Reactor Preliminary NEPA Data Study

    International Nuclear Information System (INIS)

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  17. Advanced Burner Reactor Preliminary NEPA Data Study.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  18. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

  19. Pavement Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Comprehensive Environmental and Structural Analyses The ERDC Pavement Testing Facility, located on the ERDC Vicksburg campus, was originally constructed to provide...

  20. Massachusetts Large Blade Test Facility Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Rahul Yarala; Rob Priore

    2011-09-02

    Project Objective: The Massachusetts Clean Energy Center (CEC) will design, construct, and ultimately have responsibility for the operation of the Large Wind Turbine Blade Test Facility, which is an advanced blade testing facility capable of testing wind turbine blades up to at least 90 meters in length on three test stands. Background: Wind turbine blade testing is required to meet international design standards, and is a critical factor in maintaining high levels of reliability and mitigating the technical and financial risk of deploying massproduced wind turbine models. Testing is also needed to identify specific blade design issues that may contribute to reduced wind turbine reliability and performance. Testing is also required to optimize aerodynamics, structural performance, encourage new technologies and materials development making wind even more competitive. The objective of this project is to accelerate the design and construction of a large wind blade testing facility capable of testing blades with minimum queue times at a reasonable cost. This testing facility will encourage and provide the opportunity for the U.S wind industry to conduct more rigorous testing of blades to improve wind turbine reliability.

  1. Operating manual for the critical experiments facility

    International Nuclear Information System (INIS)

    The operation of the Critical Experiments Facility (CEF) requires careful attention to procedures in order that all safety precautions are observed. Since an accident could release large amounts of radioactivity, careful operation and strict enforcement of procedures are necessary. To provide for safe operation, detailed procedures have been written for all phases of the operation of this facility. The CEF operating procedures are not to be construed to constitute a part ofthe Technical Specifications. In the event of any discrepancy between the information given herein and the Technical Specifications, limits set forth in the Technical Specifications apply. All normal and most emergency operation conditions are covered by procedures presented in this manual. These procedures are designed to be followed by the operating personnel. Strict adherence to these procedures is expected for the following reasons. (1) To provide a standard, safe method of performing all operations, the procedures were written by reactor engineers experienced in supervising the operation of reactors and were reviewed by an organization with over 30 years of reactor operating experience. (2) To have an up-to-date description of operating techniques available at all times for reference and review, it is necessary that the procedures be written

  2. Methods/data testing for advanced LWR fuel designs and operational modes in the PROTEUS facility

    International Nuclear Information System (INIS)

    In the framework of the LWR-PROTEUS programme, integral experiments are being performed for the validation of reactor physics methods/data applied to the analysis of modern light water (LWR) fuel assembly designs. One of the important aspects pertaining to this programme is the investigation of the sensitivity of integral quantities, such as reaction rate spatial distributions, local reaction rate ratios and reactivity effects, to different nuclear data sets, e.g. ENDF/B-V, ENDF/B-VI and JEF-2.2. Data sensitivity studies are currently reported, largely based on continuous-energy Monte Carlo calculations at PROTEUS-core, BWR fuel assembly and pin-cell levels. Data effects could thus be investigated systematically, independent of the inherent limitations of deterministic codes such as geometrical simplification and resonance shielding approximations. The results show a significant sensitivity of system reactivity to the data set employed, calculated values being larger by several hundreds of pcm with JEF-2.2 than with ENDF/B-V or B-VI. As regards detailed reaction rate distributions in BWR assemblies (investigated in LWR-PROTEUS Phase I), there appears to be only a weak dependence on the data library used. In the context of the investigations planned with highly burnt fuel rod segments in the Phase II experiments, certain studies made at pin-cell level have indicated that the effects of basic data differences are likely to be much lower than uncertainties in fuel composition predictions. (author)

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

    International Nuclear Information System (INIS)

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

  4. Removal heat extraction systems in advanced reactors

    International Nuclear Information System (INIS)

    The two main problems generally attributed to the electricity generation by nuclear power are the security of the facility and the radioactivity of the nuclear wastes, in a way that the only tasks of the European Commission on this matter are to make sure a high level of security in the facilities, as well as an adequate fuel and waste management. In this paper we discuss about the main lines in which the CIEMAT and the Polytechnic University of Valencia are working relative to the study of the passive working systems of the advanced designs reactors. (Author) 24 refs

  5. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  6. As-Run Physics Analysis for the UCSB-1 Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Joseph Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The University of California Santa Barbara (UCSB) -1 experiment was irradiated in the A-10 position of the ATR. The experiment was irradiated during cycles 145A, 145B, 146A, and 146B. Capsule 6A was removed from the test train following Cycle 145A and replaced with Capsule 6B. This report documents the as-run physics analysis in support of Post-Irradiation Examination (PIE) of the test. This report documents the as-run fluence and displacements per atom (DPA) for each capsule of the experiment based on as-run operating history of the ATR. Average as-run heating rates for each capsule are also presented in this report to support the thermal analysis.

  7. Experimental tests and qualification of analytical methods to address thermohydraulic phenomena in advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    Worldwide there is considerable experience in nuclear power technology, especially in water cooled reactor technology. Of the operating plants, in September 1998, 346 were light water reactors (LWRs) totalling 306 GW(e) and 29 were heavy water reactors (HWRs) totalling 15 GW(e). The accumulated experience and lessons learned from these plants are being incorporated into new advanced reactor designs. Utility requirements documents have been formulated to guide these design activities by incorporating this experience, and results from research and development programmes, with the aim of reducing costs and licensing uncertainties by establishing the technical bases for the new designs. Common goals for advanced designs are high availability, user-friendly features, competitive economics and compliance with internationally recognized safety objectives. Large water cooled reactors with power outputs of 1300 MW(e) and above, which possess inherent safety characteristics (e.g. negative Doppler moderator temperature coefficients, and negative moderator void coefficient) and incorporate proven, active engineered systems to accomplish safety functions are being developed. Other designs with power outputs from, for example, 220 MW(e) up to about 1300 MW(e) which also possess inherent safety characteristics and which place more emphasis on utilization of passive safety systems are being developed. Passive systems are based on natural forces and phenomena such as natural convection and gravity, making safety functions less dependent on active systems and components like pumps and diesel generators. In some cases, further experimental tests for the thermohydraulic conditions of interest in advanced designs can provide improved understanding of the phenomena. Further, analytical methods to predict reactor thermohydraulic behaviour can be qualified for use by comparison with the experimental results. These activities should ultimately result in more economical designs. The

  8. Developing the MAPLE materials test reactor concept

    International Nuclear Information System (INIS)

    MAPLE-MTR is a new multipurpose research facility being planned by AECL Research as a possible replacement for the 35-year-old NRU reactor. In developing the MAPLE-MTR concept, AECL is starting from the recent design and licensing experience with the MAPLE-X10 reactor. By starting from technology developed to support the MAPLE-X10 design and adapting it to produce a concept that satisfies the requirements of fuel channel materials testing and fuel irradiation programs, AECL expects to minimize the need for major advances in nuclear technology (e.g., fuel, heat transfer). Formulation of the MAPLE-MTR concept is at an early stage. This report describes the irradiation requirements of the research areas, how these needs are translated into design criteria for the project and elements of the preliminary design concept

  9. Shear compression testing of glass-fibre steel specimens after 4K reactor irradiation: Present status and facility upgrade

    International Nuclear Information System (INIS)

    The shear strengths of various fibre reinforced resins being promising candidate insulators for superconducting coils to be used tinder a strong radiation load, e.g. in future fusion reactors were investigated prior and subsequent to reactor in-core irradiation at liquid helium temperature. A large number of sandwich-like (steel-bonded insulation-steel) specimens representing a widespread variety of materials and preparation techniques was exposed to irradiation doses of up to 5 x 107 Gy in form of fast neutrons and γ-radiation. In a systematic study several experimental parameters including irradiation dose, postirradiation storage temperature and measuring temperature were varied before the determination of the ultimate shear strength. The results obtained from the different tested materials are compared. In addition an upgrade of the in-situ test rig installed at the Munich research reactor is presented, which allows combined shear/compression loading of low temperature irradiated specimens and provides a doubling of the testing rate

  10. Startup operational tests of fast reactors

    International Nuclear Information System (INIS)

    This paper is mainly concerned with the experiences of the two main phases of startup operational tests of fast reactors: (1) The general tests and Sodium filling before core loading. (2) The core loading,approach to criticality and power build up operational tests, taking for example a large and middle demonstrating integrated-type fast reactor. (author)

  11. Criticality experiments: analysis, evaluation, and programs. 7. Program of Experimental Investigations on Critical Facilities at IPPE

    International Nuclear Information System (INIS)

    In general, the BFS-1, BFS-2, MATR, and RF-GS critical facilities remain in operation at the Institute of Physics and Power Engineering (IPPE). BFS-1 and BFS-2 are universal facilities primarily intended for use with solid component elements; however, the license allows the use of water in central sub-zones. The BFS-1 and BFS-2 facilities are used for simulation of fast systems as well as systems with intermediate and thermal spectra. The MATR critical facility is intended for research of temperature effects on water-moderated cores over a wide range of temperatures (up to 250 deg. C) and pressures. The RF-GS critical facility is designed (with adaptable opportunities) for the simulation of multiplying systems with a water moderator. The experimental programs are being developed in the following directions: 1. use of plutonium released from the military programs as fuel for the BN and VVER reactors as well as for its immobilization in geological formations; 2. based on the perspective of the nuclear fuel cycle (in accordance with the initiative of the President of the Russian Federation), implemented in a new generation of fast reactors; 3. investigations of prospective reactors and accelerator-driven systems and creation of benchmarks for testing of calculational codes and neutron data for nontraditional materials. A considerable part of these investigations is being carried out in cooperation with foreign laboratories (United States, Japan, France, People's Republic of China, Republic of Korea, etc.). These programs are focused in three main directions. The first is concerned with the BN-600 reactor. Investigations into the use of a hybrid zone with this reactor have been carried out since 1999 and include the BFS-62-1 core (BN-600 initial state), BFS-62-2 (uranium core, steel reflector), BFS-62-3 (hybrid core, steel reflector), and BFS-62-4 (hybrid core, radial blanket). These experiments will be finished at the beginning of 2001 and are being carried out in

  12. AGING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2004-09-10

    The purpose of this design calculation is to revise and update the previous criticality calculation for the Aging Facility (documented in BSC 2004a). This design calculation will also demonstrate and ensure that the storage and aging operations to be performed in the Aging Facility meet the criticality safety design criteria in the ''Project Design Criteria Document'' (Doraswamy 2004, Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''SNF Aging System Description Document'' (BSC [Bechtel SAIC Company] 2004f, p. 3-12). The scope of this design calculation covers the systems and processes for aging commercial spent nuclear fuel (SNF) and staging Department of Energy (DOE) SNF/High-Level Waste (HLW) prior to its placement in the final waste package (WP) (BSC 2004f, p. 1-1). Aging commercial SNF is a thermal management strategy, while staging DOE SNF/HLW will make loading of WPs more efficient (note that aging DOE SNF/HLW is not needed since these wastes are not expected to exceed the thermal limits form emplacement) (BSC 2004f, p. 1-2). The description of the changes in this revised document is as follows: (1) Include DOE SNF/HLW in addition to commercial SNF per the current ''SNF Aging System Description Document'' (BSC 2004f). (2) Update the evaluation of Category 1 and 2 event sequences for the Aging Facility as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004c, Section 7). (3) Further evaluate the design and criticality controls required for a storage/aging cask, referred to as MGR Site-specific Cask (MSC), to accommodate commercial fuel outside the content specification in the Certificate of Compliance for the existing NRC-certified storage casks. In addition, evaluate the design required for the MSC that will accommodate DOE SNF/HLW. This design calculation will achieve the objective of providing the

  13. Environmental Test Facility (ETF)

    Data.gov (United States)

    Federal Laboratory Consortium — The Environmental Test Facility (ETF) provides non-isolated shock testing for stand-alone equipment and full size cabinets under MIL-S-901D specifications. The ETF...

  14. Ballistic Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Ballistic Test Facility is comprised of two outdoor and one indoor test ranges, which are all instrumented for data acquisition and analysis. Full-size aircraft...

  15. Advanced fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2003-04-01

    The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p-{sup 6}Li and p-{sup 11}B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D-{sup 3}He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D-{sup 3}He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of {sup 3}He per a year. On the other hand, 1 million tons of {sup 3}He is estimated to be in the moon. The {sup 3}He of about 10{sup 23} kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

  16. Status and future plan of Japan materials testing reactor

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Agency (JAEA) is a light water cooling tank typed reactor. JMTR has been used for fuel and material irradiation studies for LWRs, HTGR, fusion reactor and RI production. Since the JMTR is connected with hot laboratory through the canal, re-irradiation tests can conduct easily by safety and quick transportation of irradiation samples. First criticality was achieved in March 1968, and operation was stopped from August, 2006 for the refurbishment. The reactor facilities are refurbished during four years from the beginning of FY 2007, and necessary examination and work are carrying out on schedule. The renewed and upgraded JMTR will start from FY 2011 and operate for a period of about 20 years (until around FY 2030). The usability improvement of the JMTR, such as higher reactor available factor, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussing as the preparations for re-operation. (author)

  17. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

  18. Department of Energy's Advanced Test Reactor (ATR), July 14--18, 1980: An independent on-site safety review

    International Nuclear Information System (INIS)

    The intent of this review was not to conduct a detailed in-depth audit, but rather to make a broad management assessment of ATR operations. The results of the review should only be considered as having identified trends or indications. The Team's observations and recommendations for the most part are based upon standards used for licensed reactor facility practices. These standards form the basis for many of the comments in this report. The Team believes that a uniform minimum standard of performance should be achieved in the operation of DOE reactors. In order to assure that this is accomplished, clear standards are necessary. Consistent with the past AEC and ERDA policy, the Team has used the standards of the commercial nuclear power industry. It is recognized that this approach is conservative, in that the ATR reactor has a significantly greater degree of inherent safety (lower pressure, temperature, power, etc.) than a licensed reactor. Although the Review Team found no indications or evidence that the plant is being operated in an unsafe manner, various areas were identified where improvements are either needed or should be considered to increase the safety of reactor operations

  19. Neutronic computational modeling of the ASTRA critical facility using MCNPX

    International Nuclear Information System (INIS)

    The Pebble Bed Very High Temperature Reactor is considered as a prominent candidate among Generation IV nuclear energy systems. Nevertheless the Pebble Bed Very High Temperature Reactor faces an important challenge due to the insufficient validation of computer codes currently available for use in its design and safety analysis. In this paper a detailed IAEA computational benchmark announced by IAEA-TECDOC-1694 in the framework of the Coordinated Research Project 'Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance' was solved in support of the Generation IV computer codes validation effort using MCNPX ver. 2.6e computational code. In the IAEA-TECDOC-1694 were summarized a set of four calculational benchmark problems performed at the ASTRA critical facility. Benchmark problems include criticality experiments, control rod worth measurements and reactivity measurements. The ASTRA Critical Facility at the Kurchatov Institute in Moscow was used to simulate the neutronic behavior of nuclear pebble bed reactors. (Author)

  20. Advanced research reactor fuel development

    International Nuclear Information System (INIS)

    The fabrication technology of the U3Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U3Si2 dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U3Si2 fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 ∼ 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The γ-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U3Si2. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49

  1. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    -plates will be conducted in the Advanced Test Reactor(ATR). 49 compacts with a uranium density of 8 gU/cc consist of 7 different atomized uranium-molybdenum alloy powders. The tensile strength increased and the elongation decreased with increasing the volume fraction of U-10Mo powders in dispersion fuel. The tensile strength was lower and elongation was larger in dispersion fuel using atomized U-10Mo powders than that using comminuted fuel powders. The green strength of the comminuted powder compacts was about twice as large as that of the atomized powder compacts. It is suggested that the compacting condition required to fabricate the atomized powder compacts is over the 350MPa. The comminuted irregular shaped particles and smaller particle size of fuel powders showed improved homogeneity of powder mixture. The homogeneity of powder mixtures increased to a minimum at approximately 0.10 wt% moisture and then decreased with moisture content.

  2. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  3. Advances in heavy water reactors

    International Nuclear Information System (INIS)

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

  4. 2010 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond

    Energy Technology Data Exchange (ETDEWEB)

    mike lewis

    2011-02-01

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2009 through October 31, 2010. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of compliance activities • Discussion of the facility’s environmental impacts During the 2010 permit year, approximately 164 million gallons of wastewater were discharged to the Cold Waste Pond. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

  5. 2013 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond

    Energy Technology Data Exchange (ETDEWEB)

    Mike Lewis

    2014-02-01

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2012–October 31, 2013. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of compliance activities • Noncompliance issues • Discussion of the facility’s environmental impacts. During the 2013 permit year, approximately 238 million gallons of wastewater was discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters are below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

  6. 2012 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond

    Energy Technology Data Exchange (ETDEWEB)

    Mike Lewis

    2013-02-01

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2011 through October 31, 2012. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance issues Discussion of the facility’s environmental impacts During the 2012 permit year, approximately 183 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

  7. 2014 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Mike [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2013–October 31, 2014. The report contains the following information; Facility and system description; Permit required effluent monitoring data and loading rates; Permit required groundwater monitoring data; Status of compliance activities; Noncompliance issues; and Discussion of the facility’s environmental impacts. During the 2014 permit year, approximately 238 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters are below the Ground Water Quality Rule Secondary Constituent Standards in the downgradient monitoring wells.

  8. PROTEUS investigations for advanced thermal, fast, and intermediate-spectrum reactors

    International Nuclear Information System (INIS)

    The zero-power reactor, PROTEUS, has been used over the years for physics investigations concerning various types of advanced systems, namely gas-cooled fast reactors (GCFRs) in the seventies, light water high-conversion reactors (LWHCR) in the eighties and, currently, low-enriched-uranium high-temperature reactors (LEU-HTRs). The wide range of test neutron spectra cover underlines the versatility of the facility, the safety and operational limits for which have largely remained unaltered during the different experimental programs. This paper reviews the scope of the various PROTEUS investigations and includes evaluations of some of the most recent experiments

  9. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

  10. Design and present status of high-temperature engineering test reactor

    International Nuclear Information System (INIS)

    The Japan Atomic Energy commission (JAEC) decided to construct the high-Temperature engineering Test Reactor (HTTR) in 1987 for establishing and upgrading the basic technologies for advanced HTGRs and serving an irradiation test facility for research in high temperature technologies. The HTTR is a graphite-moderated and helium-gas-cooled test reactor with thermal output of 30MW and inlet and maximum outlet coolant temperature of 395 C and 950 C respectively. Construction started in March 1991 at Oarai site of the Japan Atomic Energy Research Institute (JAERI), with its first criticality at the end of 1997 to be followed after a series of functional tests of half a year. Fabrication of reactor pressure vessel, an intermediate heat exchanger (IHX), gas circulators and other main cooling components has been finished in their factories and installed to the site in 1994. At present, the construction of HTTR reactor building and installation of containment vessel, main and auxiliary cooling systems, etc. are almost completed. This paper describes design of the HTTR reactor cooling system, control system and present status of the HTTR construction

  11. Advanced Polymer Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Muenchausen, Ross E. [Los Alamos National Laboratory

    2012-07-25

    Some conclusions of this presentation are: (1) Radiation-assisted nanotechnology applications will continue to grow; (2) The APPF will provide a unique focus for radiolytic processing of nanomaterials in support of DOE-DP, other DOE and advanced manufacturing initiatives; (3) {gamma}, X-ray, e-beam and ion beam processing will increasingly be applied for 'green' manufacturing of nanomaterials and nanocomposites; and (4) Biomedical science and engineering may ultimately be the biggest application area for radiation-assisted nanotechnology development.

  12. Hot Fuel Examination Facility's neutron radiography reactor

    International Nuclear Information System (INIS)

    Argonne National Laboratory-West is located near Idaho Falls, Idaho, and is operated by the University of Chicago for the United States Department of Energy in support of the Liquid Metal Fast Breeder Reactor Program, LMFBR. The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both nondestructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the nondestructive examination techniques utilized at HFEF is neutron radiography, which is provided by the NRAD reactor facility (a TRIGA type reactor) below the HFEF hot cell

  13. FLUX SENSOR EVALUATIONS AT THE ATR CRITICAL FACILITY

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) and the ATR Critical (ATRC) facilities lack real-time methods for detecting thermal neutron flux and fission reaction rates for irradiation capsules. Direct measurements of the actual power deposited into a test are now possible without resorting to complicated correction factors. In addition, it is possible to directly measure minor actinide fission reaction rates and to provide time-dependent monitoring of the fission reaction rate or fast/thermal flux during transient testing. A joint Idaho State University/Idaho National Laboratory ATR National Scientific User Facility (ATR NSUF) project was recently initiated to evaluate new real-time state-of-the-art in-pile flux detection sensors. Initially, the project is comparing the accuracy, response time, and long duration performance of French Atomic Energy Commission (CEA)-developed miniature fission chambers, specialized self-powered neutron detectors (SPNDs) by the Argentinean National Energy Commission (CNEA), specially developed commercial SPNDs, and back-to-back fission (BTB) chambers developed by Argonne National Laboratory (ANL). As discussed in this paper, specialized fixturing and software was developed by INL to facilitate these joint ISU/INL evaluations. Calculations were performed by ISU 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. Ultimately, project results will be used to select the detector that can provide the best online regional ATRC power measurement. It is anticipated that project results may offer the potential to increase the ATRC's current power limit and its ability to perform low-level irradiation experiments. In addition, results from this effort will provide insights about the viability of using these detectors in the ATR. Hence, this effort complements current activities to improve ATR software tools, computational protocols

  14. Filling the gaps in SCWR materials research: advanced nuclear corrosion research facilities in Hamilton

    International Nuclear Information System (INIS)

    Research efforts on materials selection and development in support of the design of supercritical water-cooled reactors (SCWRs) have produced a considerable amount of data on corrosion, creep and other related properties. Summaries of the data on corrosion [1] and stress corrosion cracking [2] have recently been produced. As research on the SCWR advances, gaps and limitations in the published data are being identified. In terms of corrosion properties, these gaps can be seen in several areas, including: 1) the test environment, 2) the physical and chemical severity of the tests conducted as compared with likely reactor service/operating conditions, and 3) the test methods used. While some of these gaps can be filled readily using existing facilities, others require the availability of advanced test facilities for specific tests and assessments. In this paper, highlights of the new materials research facilities jointly established in Hamilton by CANMET Materials Technology Laboratory and McMaster University are presented. (author)

  15. Scaling, experiment, and code assessment on an integral testing facility

    Energy Technology Data Exchange (ETDEWEB)

    Yang, J.; Choi, S.W.; Lim, J.; Lee, D.Y.; Rassame, S.; Hibiki, T.; Ishii, M. [Purdue Univ., West Lafayette, Indiana (United States)

    2011-07-01

    A series of integral tests simulating different types of Loss-Of-Coolant Accidents (LOCAs) for new Boiling Water Reactor (BWR) design were conducted on an integral test facility (Purdue University Multi-Dimensional Integral Test Assembly, PUMA) facility. The PUMA facility was built with a scaling methodology addressing both the conservation principles and constitutive laws. A systemic study about the safety evaluation of the advanced passively safe BWR design has been performed with the collaboration of experiments on the scaled-down test facility and RELAP5/Mod3.3 code simulation. Various types of LOCA tests were performed, such as Main Steam Line Break (MSLB), Bottom Drain Line Break (BDLB), Gravity-Driven Line Break (GDLB), and Feed Water Line Break (FWLB). (author)

  16. Scaling, experiment, and code assessment on an integral testing facility

    International Nuclear Information System (INIS)

    A series of integral tests simulating different types of Loss-Of-Coolant Accidents (LOCAs) for new Boiling Water Reactor (BWR) design were conducted on an integral test facility (Purdue University Multi-Dimensional Integral Test Assembly, PUMA) facility. The PUMA facility was built with a scaling methodology addressing both the conservation principles and constitutive laws. A systemic study about the safety evaluation of the advanced passively safe BWR design has been performed with the collaboration of experiments on the scaled-down test facility and RELAP5/Mod3.3 code simulation. Various types of LOCA tests were performed, such as Main Steam Line Break (MSLB), Bottom Drain Line Break (BDLB), Gravity-Driven Line Break (GDLB), and Feed Water Line Break (FWLB). (author)

  17. Wind Tunnel Testing Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — NASA Ames Research Center is pleased to offer the services of our premier wind tunnel facilities that have a broad range of proven testing capabilities to customers...

  18. Advanced simulation for fast reactor design

    International Nuclear Information System (INIS)

    both experimentally and computationally derived constitutive relations. Accurate coupling methodologies and treatment of uncertainties for multi-physics systems are critical related areas of research. While in actuality there is no rigid distinction between such science-based and more calibration-based modeling approaches, this general advanced methodology is a significant departure from the traditional approach. This talk will give a broad, critical overview of advanced fast reactor modeling research in the context of its potential for 1) reducing uncertainties for existing fast reactor designs and 2) enabling the exploration of more innovative designs with reduced reliance on physical experiments. The degree to which sensitivity of the simulations to data, model, and geometric uncertainties can be quantified for science-based methods will be discussed in some detail, and their potential overall role in reactor design will be addressed. (author)

  19. Over view of post-irradiation examination facilities for fuels and materials development of fast reactor

    International Nuclear Information System (INIS)

    The hot cell complex for post-irradiation examination of the fast reactor fuels and materials was constructed and has been operated at the Oarai engineering center of Japan Nuclear Cycle Development Institute. The complex consists of three hot cell facilities. They are the Fuel Monitoring Facility (FMF), the Alpha-Gamma Facility (AGF) and the Materials Monitoring Facility (MMF). The FMF is located adjacent to the experimental fast reactor 'JOYO' and started operation in November 1978. In this facility, nondestructive examination of fuel subassemblies and other core components, in addition to some destructive examination of fuel and absorber pins, are carried out. The selected pins and materials, sectioned to the appropriate size at the FMF, are sent to the AGF and the MMF for further detailed examinations. The AGF has been operated successfully since October 1971. The functions of this facility are the physical, metallurgical and chemical examinations of irradiated plutonium-bearing fuels. The MMF was constructed at 1972 and has been operated since 1973 for the reactor materials. In this facility, various tests are conducted on core materials, structural material and control rod materials irradiated in fast reactor. Structural materials irradiated in JMTR and pressure tubes irradiated in prototype advanced thermal reactor 'Fugen' are also examined. (author)

  20. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  1. Toroid magnet test facility

    CERN Multimedia

    2002-01-01

    Because of its exceptional size, it was not feasible to assemble and test the Barrel Toroid - made of eight coils - as an integrated toroid on the surface, prior to its final installation underground in LHC interaction point 1. It was therefore decided to test these eight coils individually in a dedicated test facility.

  2. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  3. Preliminary concepts for materials measurement and accounting in critical facilities

    International Nuclear Information System (INIS)

    Preliminary concepts are presented for improved materials measurement and accounting in large critical facilities. These concepts will be developed as part of a study that will emphasize international safeguarding of critical facilities. The major safeguards problem is the timely verification of in-reactor inventory during periods of reactor operation. This will require a combination of measurement, statistical sampling, and data analysis techniques. Promising techniques include integral measurements of reactivity and other reactor parameters that are sensitive to the total fissile inventory, and nondestructive assay measurements of the fissile material in reactor fuel drawers and vault storage canisters coupled with statistical sampling plans tailored for the specific application. The effectiveness of proposed measurement and accounting strategies will be evaluated during the study

  4. Systemization of Design and Analysis Technology for Advanced Reactor

    International Nuclear Information System (INIS)

    The present study is performed to establish the base for the license application of the original technology by systemization and enhancement of the technology that is indispensable for the design and analysis of the advanced reactors including integral reactors. Technical reports and topical reports are prepared for this purpose on some important design/analysis methodology; design and analysis computer programs, structural integrity evaluation of main components and structures, digital I and C systems and man-machine interface design. PPS design concept is complemented reflecting typical safety analysis results. And test plans and requirements are developed for the verification of the advanced reactor technology. Moreover, studies are performed to draw up plans to apply to current or advanced power reactors the original technologies or base technologies such as patents, computer programs, test results, design concepts of the systems and components of the advanced reactors. Finally, pending issues are studied of the advanced reactors to improve the economics and technology realization

  5. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  6. Status of the IAEA coordinated research project on heat transfer behavior and thermo-hydraulics code testing for super critical water cooled reactors

    International Nuclear Information System (INIS)

    One of the key roles of the IAEA is to foster the collaboration among Member States on the development of advances in technology for advanced nuclear power plants. There is high international interest, both in developing and industrialized countries, in innovative supercritical water-cooled reactors (SCWRs), primarily because such concepts will achieve high thermal efficiencies (44-45%) and promise improved economic competitiveness utilizing and building upon the recent developments for highly efficient fossil power plants. The SCWR has been selected as one of the promising concepts for development by the Generation-IV International Forum. Following the advice of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and TWG-HWR), with the feedback from the Gen-IV SCWR Steering Committee, and in coordination with the OECD-NEA, IAEA has recently started a Coordinated Research Project (CRP) in the areas of heat transfer behaviour and testing of thermo-hydraulic computer methods for Supercritical Water-Cooled Reactors. The first Research Coordination Meeting (RCM) of the CRP was held at the IAEA Headquarters, in Vienna, Austria in July 2008. This paper summarizes the current status of the CRP, including the Integrated Research Plan and the general schedule for the CRP. (author)

  7. Advanced Neutron Source (ANS) Project progress report

    International Nuclear Information System (INIS)

    This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I ampersand C research and development; facility concepts; design; and safety

  8. Advanced Neutron Source (ANS) Project progress report

    Energy Technology Data Exchange (ETDEWEB)

    McBee, M.R.; Chance, C.M. (eds.) (Oak Ridge National Lab., TN (USA)); Selby, D.L.; Harrington, R.M.; Peretz, F.J. (Oak Ridge National Lab., TN (USA))

    1990-04-01

    This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I C research and development; facility concepts; design; and safety.

  9. Overview of irradiation facilities and experiments currently in the Oak Ridge High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    The Oak Ridge High Flux Isotope Reactor (HFIR) is an 85 MW research reactor with a variety of irradiation facilities. The target region has the highest continuous thermal neutron flux available in the western world and facilities in the beryllium reflector provide opportunities to irradiate experiments of various sizes in a variety of neutron spectrums. Major programs utilizing these facilities include Fusion Materials, Advanced Neutron Source (ANS), New Production Reactor, and Modular High Temperature Gas-Cooled Reactor

  10. Construction and commissioning test report of the CEDM test facility

    International Nuclear Information System (INIS)

    The test facility for performance verification of the control element drive mechanism (CEDM) of next generation power plant was installed at the site of KAERI. The CEDM was featured a mechanism consisting of complicated mechanical parts and electromagnetic control system. Thus, a new CEDM design should go through performance verification tests prior to it's application in a reactor. The test facility can simulate the reactor operating conditions such as temperature, pressure and water quality and is equipped with a test chamber to accomodate a CEDM as installed in the power plant. This test facility can be used for the following tests; endurance test, coil cooling test, power measurement and reactivity rod drop test. The commissioning tests for the test facility were performed up to the CEDM test conditions of 320 C and 150 bar, and required water chemistry was obtained by operating the on-line water treatment system

  11. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  12. Technical specifications for the Pajarito Site Critical Experiments Facility

    International Nuclear Information System (INIS)

    This document is to satisfy the requirement for technical specifications spelled out in DOE Manual Chapter 0540, Safety of DOE-Owned Reactors. Technical specifications are defined in Sec. 0540-048, and the requirement for them appears in Sec. 0540-015. The following technical specifications update the document, Technical Specifications for the Pajarito Site Critical Experiments Facility

  13. Design study on the Advanced Recycling Reactor

    International Nuclear Information System (INIS)

    Full text: The design study on the Advanced Recycling Reactor (ARR) has been conducted. This paper presents the pre-conceptual design of the ARR that is a loop-typed sodium cooled reactor with MOX fuel. International Nuclear Recycling Alliance (INRA) takes advantage of international experience and uses the design based on Japan Sodium-cooled Fast Reactor (JSFR) as reference for FOA studies of US DOE, because Japan has conducted R and Ds for the JSFR incorporating thirteen technology enhancements expected to improve safety, enhance economics, and increase reactor reliability. The targets of the ARR are to generate electricity while consuming fuel containing transuranics and to attain cost competitiveness with the similar sized LWRs. INRA proposes 3 evolutions of the ARR; ARR1, a 500 MWe demonstration plant, online in 2025; ARR2, a 1,000 MWe commercial plant, online in 2035; ARR3, a 1,500 MWe full-scale commercial plant, online in 2050. INRA believes the scale-up factor of two is acceptable increase from manufacturing and licensing points of view. Major features of the ARR1 are the following: The reactor core is 70cm high and the volume fraction of fuel is approximately 32%. The conversion ratio of fissile is set up less than 0.6 and the amount of burned TRU is 45-51 kg/TWeh.Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop arrangement and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The ARR1 is co-located with a recycling facility. The overall plant facility arrangement is planned assuming to be constructed and installed in an inland area. The plant consists of a reactor building (including reactor auxiliary facilities and electrical/control systems), a turbine building, and a recycling building. The volume of the reactor building will be approximately 180,000 m3. The capital cost for the ARR1 and the ARR2 are

  14. Preventive and Predictive Maintenance, Warehousing of Spares, Periodic Testing and In-Service Inspection Activities at the Nigerian Research Reactor-1 Facility

    International Nuclear Information System (INIS)

    The Nigerian Research Reactor–1, or NIRR-1, is sited at Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. Activities on preventive or routine maintenance have been institutionalized since the commissioning of the reactor in February 2004. This has grossly reduced the rates of corrective maintenance activities and helped the reactor management a great deal in predicting failure rates of reactor components and other auxiliary units. Routine maintenance of systems and components are being carried out on a weekly, quarterly and annual basis based on manufacturer’s recommendations, which have been reviewed and improved over the years. The paper presents the implementation of maintenance activities in NIRR-1 from its initial criticality in 2004 till today and the new scheme for periodic testing and in-service-inspection developed after an IAEA Integrated Safety Assessment of Research Reactors mission. The measures put in place are envisaged to reduce the negative impact of ageing on NIRR-1 and its auxiliary systems. (author)

  15. Testing the reactor charging machine

    International Nuclear Information System (INIS)

    One of the main objective of the R - D technological engineering program devoted to the Fuel Handling System is domestic production of equipment and technology for testing the ends of the reactor charging machine (MID) destined to Cernavoda NPP, beginning with Unit 2. To achieve the objective based on an own design, a bench-scale testing stand of MIDs which can simulate the pressure, flow-rate, and temperature conditions proper to fuel channels in operating CANDU 600 reactors. The main components of this testing facility are: - fuel channels, cold also test sections, allowing the coupling of MID end upwardly and downwardly, corresponding to the direction of the water flow through the channel; - technological installation feeding with light water the testing sections of the facility in thermohydraulic conditions, similar to those in the reactor, allowing the cold and hot testings, respectively, of the MID end; - cold testing installation, water supply and oil control panel, feeding the hydraulic drives of the MID's end during the testings; - fixed bridge and mobile carrier for MID's end positioning against testing sections; - installation for functional testing of MID thrusters, before pre-admission and reception tests; - dedicated tools and devices; - raising and transport mechanical devices for handling and positioning the MID's end upon the carrier; - automation panel for controlling the stand equipment and MID's end; - process computer for conducting on-line tests. MID's end testing implies mainly the following operations: - regulation, calibration and functional testing of the MID thrusters carried out independently on a specialised stand; - regulation and calibration of MID's end sub-assemblages; - carrying out the cold and hot pre-admission tests consisting in automatic performing, without operator intervention, of 12 fuel changes, two of which being successive; - performing the cold and hot reception tests, consisting in automatic accomplishment of 4

  16. Preserving physics knowledge at the fast flux test facility

    International Nuclear Information System (INIS)

    One of the goals of the Dept. of Energy's Office of Nuclear Energy, initiated under the Fuel Cycle Research and Development Program (FCRD) and continued under the Advanced Reactor Concepts Program (ARC) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMRs) that could support the development of an environmentally and economically sound nuclear fuel cycle. The Fast Flux Test Facility (FFTF) is the most recent LMR to operate in the United States, from 1982 to 1992, and was designed as a fully instrumented test reactor with on-line, real time test control and performance monitoring of components and tests installed in the reactor. The 10 years of operation of the FFTF provided a very useful framework for testing the advances in LMR safety technology based on passive safety features that may be of increased importance to new designs after the events at Fukushima. Knowledge preservation at the FFTF is focused on the areas of design, construction, and startup of the reactor, as well as on preserving information obtained from 10 years of successful operating history and extensive irradiation testing of fuels and materials. In order to ensure protection of information at risk, the program to date has sequestered reports, files, tapes, and drawings to allow for secure retrieval. A disciplined and orderly approach has been developed to respond to client's requests for documents and data in order to minimize the search effort and ensure that future requests for this information can be readily accommodated. (authors)

  17. Electromagnetic propulsion test facility

    Science.gov (United States)

    Gooder, S. T.

    1984-01-01

    A test facility for the exploration of electromagnetic propulsion concept is described. The facility is designed to accommodate electromagnetic rail accelerators of various lengths (1 to 10 meters) and to provide accelerating energies of up to 240 kiloJoules. This accelerating energy is supplied as a current pulse of hundreds of kiloAmps lasting as long as 1 millisecond. The design, installation, and operating characteristics of the pulsed energy system are discussed. The test chamber and its operation at pressures down to 1300 Pascals (10 mm of mercury) are described. Some aspects of safety (interlocking, personnel protection, and operating procedures) are included.

  18. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    Directory of Open Access Journals (Sweden)

    Sabharwall Piyush

    2015-01-01

    Full Text Available A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX and a secondary heat exchanger (SHX. Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C, high-pressure (7 MPa helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 flow loop operating at low pressure (0.2 MPa, at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift in measuring operational data for extended periods of times, as data collected will be

  19. System-integrated modular advanced reactor (SMART)

    International Nuclear Information System (INIS)

    coolant pump (MCP) and pressurizer (PZR), are installed in a single reactor vessel assembly (RVA). The conceptual and basic designs of SMART with a desalination system were completed in March of 1999 and March of 2002, respectively. SMART development has been conducted under the nuclear research and development programme supported by the Ministry of Science and Technology (MOST) of the Republic of Korea and thus KAERI and MOST are the principal stakeholders. The SMART design focuses on the enhancement of safety and improvement of the reliability as well as the economics. For these purposes, highly advanced design features enhancing the safety, reliability, performance, and operability were introduced into the SMART design. Advanced design features should be proven or qualified by experience, testing, or analysis and, if possible, the equipment should be designed according to approved standards. Some fundamental thermal-hydraulic experiments were carried out during the design concept development to assure the fundamental behaviour of major concepts of the SMART systems. Various thermal-hydraulic and mechanical tests are in progress and planned. In addition, overall SMART performance will be demonstrated through the SMART pilot plant construction and operation

  20. JAEA key facilities for global advanced fuel cycle R and D

    International Nuclear Information System (INIS)

    Advanced fuel cycle will be realized with the mid and long term R and D during the long-term transition period from LWR cycle to advanced reactor fuel cycle. Most of JAEA facilities have been utilized to establish the current LWR and FBR (Fast Breeder Reactor) fuel cycle by implementing evolutionary R and D. An assessment of today's state experimental facilities concerning the following research issues: reprocessing, Mox fuel fabrication, irradiation and post-irradiation examination, waste management and nuclear data measurement, is made. The revolutionary R and D requests new issues to be studied: the TRU multi-recycling, minor actinide recycling, the assessment of proliferation resistance and the assessment of cost reduction. To implement the revolutionary R and D for advanced fuel cycle, however, these facilities should be refurbished to install new machines and process equipment to provide more flexible testing parameters

  1. Code qualification of structural materials for AFCI advanced recycling reactors

    International Nuclear Information System (INIS)

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the

  2. Analyses of the OSU-MASLWR Experimental Test Facility

    International Nuclear Information System (INIS)

    Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well.

  3. National Solar Thermal Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The National Solar Thermal Test Facility (NSTTF) is the only test facility in the United States of its type. This unique facility provides experimental engineering...

  4. Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

    International Nuclear Information System (INIS)

    This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

  5. Advanced Recycling Reactor with Minor Actinide Fuel

    International Nuclear Information System (INIS)

    The Advanced Recycling Reactor (ARR) with minor actinide fuel has been studied. This paper presents the pre-conceptual design of the ARR proposed by the International Nuclear Recycling Alliance (INRA) for FOA study sponsored by DOE of the United States of America (U.S.). Although the basic reactor concept is technically mature, it is not suitable for commercial use due to the need to reduce capital costs. As a result of INRA's extensive experience, it is anticipated that a non-commercial ARR1 will be viable and meet U.S. requirements by 2025. Commercial Advanced Recycling Reactor (ARR) operations are expected to be feasible in competition with LWRs by 2050, based on construction of ARR2 in 2035. The ARR based on the Japan Sodium-cooled Fast Reactor (JSFR) is a loop-typed sodium cooled reactor with MOX fuel that is selected because of much experience of SFRs in the world. Major features of key technology enhancements incorporated into the ARR are the following: Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop system and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The reactor core of the ARR1 is 70 cm high and the volume fraction of fuel is 31.6%. The conversion ratio of fissile is set up less than 0.65 and the amount of burned TRU is 45-51 kg/TWeh. According to survey of more effective TRU burning core, the oxide fuel core containing high TRU (MA 15%, Pu 35% average) with moderate pins of 12% arranged driver fuel assemblies can decrease TRU conversion ratio to 0.33 and improve TRU burning capability to 67 kg/TWeh. The moderator can enhance TRU burning, while increasing the Doppler effect and reducing the positive sodium void effect. High TRU fraction promotes TRU burning by curbing plutonium production. High Am fraction and Am blanket promote Am transmutation. The ARR1 consists of a reactor building (including

  6. Future Transient Testing of Advanced Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2009-09-01

    The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by

  7. Conceptual studies of plasma engineering test facility

    International Nuclear Information System (INIS)

    Conceptual studies have been made of a Plasma Engineering Test Facility, which is to be constructed following JT-60 prior to the experimental power reactor. The physical aim of this machine is to examine self-ignition conditions. This machine possesses all essential technologies for reactor plasma, i.e. superconducting magnet, remote maintenance, shielding, blanket test modules, tritium handling. Emphasis in the conceptual studies was on structural consistency of the machine and whether the machine would be constructed practically. (author)

  8. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  9. Design and construction of a fast critical facility

    International Nuclear Information System (INIS)

    Design and construction of a fast critical facility. In a fast-power-reactor development programme, a critical facility is found to be a highly useful tool to ascertain calculational techniques, to verify neutron cross-section sets, and to obtain integral reactor-physics parameters necessary for the nuclear design of a power system. Since it is primarily a physics instrument, the design of a fast critical facility itself poses a number of different problems not found in the design of a power reactor. In addition to usual questions of site, containment, core design and instrumentation , there arise such problems as: how to obtain a large degree of flexibility consistent with safety, the determination of the size and type of facility to meet the experimental physics requirements, the determination of the number and location of control and safety rods minimizing perturbation effects and the specification of the reproducibility of control rods and other movable components to obtain the accuracy required in reactivity measurements. These are some of the problems which are discussed in this paper based on recent experience at the Argonne National Laboratory which has under construction a fast critical facility, ZPR-VI at its Lemont, Illinois site for fast-reactor-physics studies. The ZPR-VI is a movable half- or split-table-type machine similar to ZPR-III. It has a matrix about two and a half times the volume of the earlier machine and will be used to investigate the physics of large, highly dilute, metal and cermet, unmoderated and partially moderated systems having core volumes up to about 1500 l. A detailed description of the ZPR-VI with a discussion on the criteria used in the design of its various components from the point of view of reactor physics is presented. In addition, such topics as management and operating procedures, potential hazards during operation, experimental techniques to be used and construction costs are also included. (author)

  10. Flux mapping system for AHWR critical facility

    International Nuclear Information System (INIS)

    A software for flux mapping system (FMS) for AHWR critical facility has been developed. The system consists of 25 LEU based pulse detectors and associated software. The objective of the FMS is to obtain the flux profiles over the central 5 x 5 lattice locations. For development of flux mapping system it is required to compute the higher harmonics of the diffusion equation. These harmonics (also called λ-modes) are the eigen functions of multi-group diffusion equation. The fundamental mode is found by power iteration method. Apart from the fundamental mode, other higher modes are evaluated by subtraction technique. In the present paper, fundamental eigenvalue and eigenfunction are evaluated by finite difference method. The bi-orthogonality relations between the direct and adjoint eigenvectors are used for this purpose. The reference core for AHWR has been simulated by computer code FINSQR. Two group lattice cell data have been generated using transport theory code WIMSD and its associated 69-group nuclear data library. The reactor core along with the surrounding radial and axial reflector was represented using 3243 mesh points. In total 5 λ-modes and corresponding eigenvalues have been estimated. The computer code FMS has been developed specifically for AHWR/PHWR critical facility. Flux construction at 5 x 5 lattice locations of the core has been achieved by using observed fluxes at 25 detector locations and linear combinations of pre calculated eigen functions. Combining coefficients have been computed by least square method. To validate the code, we have used computer code based on Monte Carlo method for estimation of thermal fluxes at a few discrete locations. The fluxes were estimated using ENDF/B-VI point nuclear data library. (author)

  11. Power Systems Development Facility Gasification Test Run TC07

    Energy Technology Data Exchange (ETDEWEB)

    Southern Company Services

    2002-04-05

    This report discusses Test Campaign TC07 of the Kellogg Brown & Root, Inc. (KBR) Transport Reactor train with a Siemens Westinghouse Power Corporation (Siemens Westinghouse) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The Transport Reactor is an advanced circulating fluidized-bed reactor designed to operate as either a combustor or a gasifier using a particulate control device (PCD). The Transport Reactor was operated as a pressurized gasifier during TC07. Prior to TC07, the Transport Reactor was modified to allow operations as an oxygen-blown gasifier. Test Run TC07 was started on December 11, 2001, and the sand circulation tests (TC07A) were completed on December 14, 2001. The coal-feed tests (TC07B-D) were started on January 17, 2002 and completed on April 5, 2002. Due to operational difficulties with the reactor, the unit was taken offline several times. The reactor temperature was varied between 1,700 and 1,780 F at pressures from 200 to 240 psig. In TC07, 679 hours of solid circulation and 442 hours of coal feed, 398 hours with PRB coal and 44 hours with coal from the Calumet mine, and 33 hours of coke breeze feed were attained. Reactor operations were problematic due to instrumentation problems in the LMZ resulting in much higher than desired operating temperatures in the reactor. Both reactor and PCD operations were stable and the modifications to the lower part of the gasifier performed well while testing the gasifier with PRB coal feed.

  12. Completion summary for boreholes USGS 140 and USGS 141 near the Advanced Test Reactor Complex, Idaho National Laboratory, Idaho

    Science.gov (United States)

    Twining, Brian V.; Bartholomay, Roy C.; Hodges, Mary K.V.

    2014-01-01

    organic compounds, stable isotopes, and radionuclides. Water samples from both wells indicated that concentrations of tritium, sulfate, and chromium were affected by wastewater disposal practices at the Advanced Test Reactor Complex. Most constituents in water from wells USGS 140 and USGS 141 had concentrations similar to concentrations in well USGS 136, which is upgradient from wells USGS 140 and USGS 141.

  13. Advanced fuel in the Budapest research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hargitai, T.; Vidovsky, I. [KFKI Atomic Energy Research Inst., Budapest (Hungary)

    1997-07-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  14. Environmental test facilities

    International Nuclear Information System (INIS)

    This paper discusses environmental test facilities which are used to simulate an environment or combination of environments under laboratory controlled conditions that duplicate or exaggerate the effects found in actual service. They assist the engineer and scientist in exploring the effects of equipment and in developing equipment for resistance to the many environmental forces. The acceptance of and demand for environmental simulation facilities result form the following factors: parallel and reproducible tests can be made; equipment being tested can usually be observed and analyzed during testing; and supporting equipment requirements are reduced to a minimum. Field testing and product development costs are reduced, lead time required for completion of product development is shortened, and most desirable reliability features can be incorporated in the original manufacture of the product. Environmental equipment is used not only to determine the performance of mechanical and electrical equipment, but for certain tests on personnel as well. Personnel testing includes: checking protective equipment and clothing; altitude and space procedures indoctrination; and studying physiological and psychological effects on the human body and mind. Environmental testing is usually divided into two general classifications-climatic and dynamic

  15. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author).

  16. Criticality safety and facility design considerations

    International Nuclear Information System (INIS)

    Operations with fissile material introduce the risk of a criticality accident that may be lethal to nearby personnel. In addition, concerns over criticality safety can result in substantial delays and shutdown of facility operations. For these reasons, it is clear that the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The emphasis of this report will be placed on engineering design considerations in the prevention of criticality. The discussion will not include other important aspects, such as the physics of calculating limits nor criticality alarm systems

  17. Practical nuclear power training for overseas trainees using reactor facilities and radiation handling facilities

    International Nuclear Information System (INIS)

    The research reactor of Tokyo City University Atomic Energy Research Laboratory (Musashi Institute of Technology reactor) is zirconium-moderated water-cooled solid homogeneous type (TRIGA-II type), and its maximum heat output is 100 kW. It got into the first critical state in January 1963, and since then, it has achieved success in many researches. Although its decommissioning was decided in 2013, the existing facilities are used in education, and the research related to the decommissioning of research reactor facilities is carried out. Radiation handling facilities are in place, and they are widely used in education and research activities. Atomic Energy Research Laboratory, as a place for education, is conducting education and research activities such as the education using radiation handling facilities, development of an actual feeling type reactor operation simulator using the control panel of Musashi Institute of Technology reactor and operation performance data. This paper reports the practical nuclear power training for overseas trainees using the reactor facilities and radiation handling facilities. It also reports training implementation plan, acceptance preparation, contents of training, and the results of training. (A.O.)

  18. Fiscal year 1999 multi-year work plan, advanced reactors transition program

    International Nuclear Information System (INIS)

    The Advanced Reactors Transition (ART) has two missions. One, funded by DOE-EM is to transition assigned, surplus facilities to a safe and compliant, low-cost stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D and D. Facilities to be transitioned include the 309 Building/Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy (NE) Legacy Facilities. The second mission, funded by DOE-NE, is to maintain the Fast Flux Test Facility (FFTF) and affiliated 400 Area buildings in a safe and compliant standby condition. The condition of the plant hardware, software and personnel is to be preserved in a manner not to preclude a plant restart

  19. Strategic Need for Multi-Purpose Thermal Hydraulic Loop for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; Su-Jong Yoon; Gregory K. Housley

    2014-09-01

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

  20. Nondestructive testing of welds in steam generators for advanced gas cooled reactors at Heyshamm II and Torness

    International Nuclear Information System (INIS)

    The paper concerns non-destructive testing (NDT) of welds in advanced gas cooled steam generators for Heysham II and Torness nuclear power stations. A description is given of the steam generator. The selection of NDT techniques is also outlined, including the factors considered to ascertain the viability of a technique. Examples are given of applied NDT methods which match particular fabrication processes; these include: microfocus radiography, ultrasonic testing of austenitic tube butt welds, gamma-ray isotope projection system, surface crack detection, and automated radiography. Finally, future trends in this field of NDT are highlighted. (UK)

  1. The study of the test section on the fusion engineering facility

    International Nuclear Information System (INIS)

    The research and development of a nuclear fusion reactor are rapidly making progress in plasma physics, superconductive magnets, heating technology of plasma and so on. The detailed study of a large-scale fusion reactor which uses DT plasma has been started as a next stage of a critical nuclear fusion reactor (JT-60). While the technology of plasma confinement is being developed, the study of reactor engineering technology is very important to put a nuclear fusion reactor to practical use. Therefore the conceptual facility called the fusion engineering facility (FEF) is proposed to be effective for a nuclear fusion reactor development. This report describes the utilization of the test section on the FEF and the characteristics of neutron flux at the test section. (author)

  2. Progress in the Development of the Modular Pebble-Bed Advanced High Temperature Reactor

    International Nuclear Information System (INIS)

    This review article summarizes recent progress by students and faculty at U.C. Berkeley working on the development of the Pebble-Bed Advanced High Temperature Reactor (PB-AHTR). The 410-MWe PBAHTR is a liquid salt cooled reactor that operates at near atmospheric pressure and high power density (20 to 30 MW/m3, compared to 4.8 MW/m3 for helium cooled reactors). Operating with a core inlet temperature of 600 deg. C and outlet temperature of 704 deg. C, the PB-AHTR uses well understood materials of construction including Alloy 800H with Hastelloy N cladding for the reactor vessel and primary loop components, and graphite for core and reflector structures. Recent work by the NE 170 senior design class has developed physical arrangements for the major reactor and power conversion components, along with the structural design for the reactor building and turbine hall featuring seismic base isolation, design for aircraft crash protection, shielding analysis, and design of a multiple-zone ventilation and containment system to provide effective control of radioactive and chemical contamination. The resulting total building volume is 260 m3/MWe, compared to 343 m3/MWe to 486 m3/MWe for current large (1150 to 1600 MWe) LWR designs. These results suggest the potential for significant reductions in construction time and cost. Neutronics studies have verified the capability to design the PB-AHTR with negative fuel and coolant temperature reactivity coefficients, for both LEU and deep-burn TRU fuels. Depletion analysis was also performed to identify optimal core designs to maximize fuel utilization. The additional moderation provided by the coolant simplifies design to achieve optimal moderation, and the spent fuel volume is approximately half that of helium cooled reactors. In collaboration with the Czech Nuclear Research Institute, initial zero-power critical tests were performed to validate PB-AHTR neutronics models. Liquid salts are unique among candidate reactor coolants due

  3. The physics of accelerator driven sub-critical reactors

    Indian Academy of Sciences (India)

    S B Degweker; Biplab Ghosh; Anil Bajpal; S D Pranjape

    2007-02-01

    In recent years, there has been an increasing worldwide interest in accelerator driven systems (ADS) due to their perceived superior safety characteristics and their potential for burning actinides and long-lived fission products. Indian interest in ADS has an additional dimension, which is related to our planned large-scale thorium utilization for future nuclear energy generation. The physics of ADS is quite different from that of critical reactors. As such, physics studies on ADS reactors are necessary for gaining an understanding of these systems. Development of theoretical tools and experimental facilities for studying the physics of ADS reactors constitute important aspect of the ADS development program at BARC. This includes computer codes for burnup studies based on transport theory and Monte Carlo methods, codes for studying the kinetics of ADS and sub-critical facilities driven by 14 MeV neutron generators for ADS experiments and development of sub-criticality measurement methods. The paper discusses the physics issues specific to ADS reactors and presents the status of the reactor physics program and some of the ADS concepts under study.

  4. The need to address the larger universe of HEU-fueled reactors, including critical assemblies, pulsed reactors and propulsion reactors

    International Nuclear Information System (INIS)

    Full text: The RERTR program has focused thus far primarily on ending shipments of HEU fuel to research reactors. This has resulted in giving highest priority to reactors with steady thermal powers of 1 megawatt or more, because they require regular refuelling. Critical facilities and pulsed reactors can also of serious concern, because some of them contain very large amounts of barely-irradiated HEU and plutonium. They could be costly to convert - and conversion to LEU may be impractical for fast-neutron critical assemblies. An assessment should be carried out first, therefore, as to which are still needed. Critical assemblies are required today primarily to benchmark Monte Carlo neutron-transport codes. Perhaps the world nuclear community could share a few instead of each reactor-design institute having its own. There is also a whole universe of HEU-fuelled pressurized-water reactors used to power submarines and other types of nuclear-powered ships. These reactors collectively require much more HEU fuel each year than research reactors. The risk of HEU diversion from their fuel cycles is not zero but it is difficult for outsiders to discuss conversion because of the fuel designs are classified. This makes the conversion of Russia's civilian icebreaker reactors of particular interest because issues of classified fuel design are less problematic and these reactors load annually fuel containing about 400 kg of U-235. Another reason for interest in developing LEU fuel for these reactors is that the KLT-40 icebreaker reactor is being adapted for a floating nuclear power plant. Finally, the research-reactor community is, in any case, faced with developing fuels that can operate at power-reactor-fuel temperatures because there are a few high-powered research reactors that operate in this temperature range. (author)

  5. Removal of reactor cooling system facilities and others in reconstruction works of JRR-3

    International Nuclear Information System (INIS)

    The home-manufactured No.1 reactor 'JRR-3' in Tokai Research Establishment, Japan Atomic Energy Research Institute, stopped its operation in March, 1983, with the results of cumulative operation time 47,135.5 hours and cumulative power output 419,073 MWh. Since then, the reconstruction works to construct a research reactor with higher performance has been advanced, and the new reactor has attained the initial criticality on March 22, 1990. The removal of the facilities which are not used after the reconstruction was carried out since August, 1985, and in this report, the removal of the parts with relatively high dose equivalent such as reactor cooling system facilities is outlined, except the reactor proper. The range of the removed facilities, the planning of removal works, the progress of removal works and the results are reported. About 110 t of large equipments were preserved as they are, and about 400 t of pipings and others were dismantled, removed, cut and put in containers. (K.I.)

  6. Commercial Light Water Reactor Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  7. Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

    1998-10-01

    The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

  8. Environmental concerns in regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN

    International Nuclear Information System (INIS)

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the main programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel -CCN- is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt % 235U), to supply its IEA-R1 research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts.This study aims to implant the Sustainability Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to CCN

  9. Environmental concerns regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN

    International Nuclear Information System (INIS)

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the maim programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel - CCN - is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt% 235U), to supply its IEA-RI research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the Sustainable Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to CCN

  10. Administrative Aspects of the Criticality Controls Used in Programmes for Basic Criticality Research, Reactor Development and Materials Processing

    International Nuclear Information System (INIS)

    This paper describes the administrative and procedural aspects of criticality controls used by a field office of the United States Atomic Energy Commission in programmes that include reactor criticals, research and materials testing reactors, and power reactor development. Situations encountered include handling, storing, and processing large quantities of uranium-235 and plutonium-239 of various configurations and compositions in laboratories and operations which gather basic criticality data, processing of fissile material, and varied reactor research and development, programmes including fuel materials. Similar situations exist for uranium-233 and plutonium-238 on a smaller laboratory scale. The administrative controls and interactions of the USAEC field office and the operating contractors, who operate these installations for the USAEC, are outlined. Also, the purpose and scope of the direct examination by USAEC personnel of these contractor facilities are analysed. The programme has been in effect for three years and is believed to be successful in maintaining efficient operations and an acceptable low level of risk of inadvertent criticality. Success of this programme is in good measure due to the close working relationship between the staffs of the USAEC field office and the operating contractors. (author)

  11. An experiment to test advanced materials impacted by intense proton pulses at CERN HiRadMat facility

    CERN Document Server

    Bertarelli, A; Boccone, V; Carra, F; Cerutti, F; Charitonidis, N; Charrondiere, C; Dallocchio, A; Fernandez Carmona, P; Francon, P; Gentini, L; Guinchard, M; Mariani, N; Masi, A; Marques dos Santos, S D; Moyret, P; Peroni, L; Redaelli, S; Scapin, M

    2013-01-01

    Predicting the consequences of highly energetic particle beams impacting protection devices as collimators or high power target stations is a fundamental issue in the design of state-of-the-art facilities for high-energy particle physics. These complex dynamic phenomena can be successfully simulated resorting to highly non-linear numerical tools (Hydrocodes). In order to produce accurate results, however, these codes require reliable material constitutive models that, at the extreme conditions induced by a destructive beam impact, are scarce and often inaccurate. In order to derive or validate such models a comprehensive, first-of-its-kind experiment has been recently carried out at CERN HiRadMat facility: performed tests entailed the controlled impact of intense and energetic proton pulses on a number of specimens made of six different materials. Experimental data were acquired relying on embedded instrumentation (strain gauges, temperature probes and vacuum sensors) and on remote-acquisition devices (laser ...

  12. The Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    The U. S. Advanced Light Water Reactor Program is a forward-looking program designed to produce viable nuclear generating system candidates to meet the very real, and perhaps imminent, need for new power generation capacity in the U. S. and around the world. The ALRR Program is an opportunity to move ahead with confidence, to confront problems today which must be confronted if the U. S. electrical utilities are to continue to meet their commitment to provide safe, reliable, economical electrical power to the nation in the years ahead. Light water reactor technology is today playing a vital role in the production of electricity to meet the world's needs. At present about 13% of the world's electricity is supplied by nuclear power plants, most of those light water reactors. Nevertheless, there is a clear need for expanded use of nuclear generation. Here in Korea and elsewhere in Asia, demand for electricity has continued to increase at a very high rate. In the United States demand growth has been more moderate, but a large number of existing stations will be ready for replacement in the next two decades, and all countries face the problem of dwindling fuel supplies and growing environmental impact of fossil-fired power plants. Despite the evident need for expanded nuclear generation capacity in the United States, there have been no new plants ordered in the past ten years and at present there are no immediate prospects for new plant orders. Concerns about safety, the high cost of recent nuclear stations, and the current excess of electrical generation capacity in the United States, have combined to interrupt completely the growth of this vital power supply system

  13. Risk analysis of the aqueous fast reactor fuel cycle facility in the conceptual design stage

    International Nuclear Information System (INIS)

    This paper describes the radioactive release risk of the advanced aqueous reprocessing and fabrication facility for the fast reactor fuel cycle. Because this advanced facility is still in the conceptual design stage, the risk analysis aims at grasping the entire risk comprehensively as simple as possible. As a potential hazard, it was shown that the main process in the reprocessing and fuel fabrication facilities involved only an order of 10-3 of radioactivity in the single reactor core of large scale. Abnormal phenomena related to radioactive solution that can cause radioactive release from the facility to the environmental atmosphere in a large quantity were identified as follows: in-vessel boiling caused by loss of cooling system, a leak and fire of inflammable organic solvent in a cell, in-vessel boiling due to criticality accident, an explosion. Simplified estimation about the quantitative risk of radioactive release showed that in-vessel boiling due to loss of cooling system had the largest contribution to the non-volatile radioactive substance release in a large quantity and that criticality accidents initiated from incomplete extraction stripping of Pu nuclides were dominant in the release risk of radioactive iodine and noble gas with a short-half-life. (author)

  14. 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

    International Nuclear Information System (INIS)

    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) began 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

  15. 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

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    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) began 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

  16. Liquid Metal Fast Breeder Reactor Program (LMFBR): facility profiles

    International Nuclear Information System (INIS)

    A description is presented of the experimental test facilities involved in the conduct of the LMFBR research and development program. Existing facilities and those under construction or authorized as of October 1975 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. Introductory material for each section includes site and facility maps and an alphabetical list of the profiles contained in the section. A glossary of acronyms and letter designations in common usage to describe organizations, reactor and test facilities, components, etc. involved in the LMFBR program is included. Alphabetical, organizational, and programmatic indexes are provided as a convenient method of identifying the facilities with their locations and with their principal uses in the LMFBR program

  17. Reactor Simulator Testing Overview

    Science.gov (United States)

    Schoenfeld, Michael P.

    2013-01-01

    Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  18. Present status of decommissioning in the Musashi Reactor Facility

    International Nuclear Information System (INIS)

    The Musashi reactor facility (TRIGA-II type) had been used as the place of studies such as the education and training of nuclear engineers, research on an activation analysis or reactor physics and a university joint research use mainly on medical and bio therapy study since the first criticality in 1963. Unfortunately a coolant leak was found from a core tank in December 1989 and stopped the operation of the reactor since. The Musashi reactor decommissioning was decided in May 2003 and took a step to shut down the reactor permanently. In 2006 all the spent fuels were returned to U.S.A. Outlines of the transport works were presented as present status of the decommissioning. (T. Tanaka)

  19. Advanced gas-cooled reactors (AGR)

    International Nuclear Information System (INIS)

    The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given

  20. Advances of study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R and D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods, grid spacer configuration etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.0 mm) and the experimental data reveal the feasibility of RMWR. (authors)

  1. Functional and performance evaluation of 28 bar hot shutdown passive valve (HSPV) at integral test loop (ITL) for advanced heavy water reactor (AHWR)

    International Nuclear Information System (INIS)

    During reactor shutdown in advanced heavy water reactor (AHWR), core decay heat is removed by eight isolation condensers (IC) submerged in gravity driven water pool. Passive valves are provided on the down stream of each isolation condenser. On increase in steam drum pressure beyond a set value, these passive valves start opening and establish steam flow by natural circulation between the four steam drums and corresponding isolation condensers under hot shutdown and therefore they are termed as Hot Shut Down Passive Valves (HSPVs). The HSPV is a self acting type valve requiring no external energy, i.e. neither air nor electric supply for actuation. This feature makes the valve functioning independent of external systems such as compressed air supply or electric power supply, thereby providing inherent safety feature in line with reactor design philosophy. The high pressure and high temperature HSPV s for nuclear reactor use, are non-standard valves and therefore not manufactured by the valve industry worldwide. In the process of design and development of a prototype valve for AHWR, a 28 bar HSPV was configured and successfully tested at Integral Test Loop (ITL) at Engineering Hall No.7. During ten continuous experiments spread over 14 days, the HSPV has proved its functional capabilities and its intended use in decay heat removal system. The in-situ pressure setting and calibration aspect of HSPV has also been successfully established during these experiments. This report gives an insight into the HSPV's functional behavior and role in reactor decay heat removal system. The report not only provides the quantitative measure of performance for 28 bar HSPV in terms of valve characteristics, pressure controllability, linearity and hysteresis but also sets qualitative indicators for prototype 80 bar HSPV, being developed for AHWR. (author)

  2. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  3. Recent Advances in Antenna Measurement Techniques at the DTU-ESA Spherical Near-Field Antenna Test Facility

    DEFF Research Database (Denmark)

    Breinbjerg, Olav; Pivnenko, Sergey; Kim, Oleksiy S.; Nielsen, Jeppe Majlund

    2014-01-01

    This paper reports recent antenna measurement projects and research at the DTU-ESA Spherical Near-Field Antenna Test Facility at the Technical University of Denmark. High-accuracy measurement projects for the SMOS, SENTINEL-1, and BIOMASS missions of the European Space Agency were driven by...... uncertainty requirements of a few hundredths of dB for the directivity and correspondingly strong requirements for gain and/or phase. Research and development of 1:3 bandwidth range probes, and the near-field to far-field transformation algorithm accounting for the higher-order azimuthal modes in the...

  4. An experiment to test advanced materials impacted by intense proton pulses at CERN HiRadMat facility

    Energy Technology Data Exchange (ETDEWEB)

    Bertarelli, A., E-mail: alessandro.bertarelli@cern.ch [CERN, Engineering Department, Mechanical and Materials Engineering Group (EN-MME), CH-1211 Geneva 23 (Switzerland); Berthome, E. [CERN, Engineering Department, Mechanical and Materials Engineering Group (EN-MME), CH-1211 Geneva 23 (Switzerland); Boccone, V. [CERN, Engineering Department, Sources, Targets and Interactions Group (EN-STI), CH-1211 Geneva 23 (Switzerland); Carra, F. [CERN, Engineering Department, Mechanical and Materials Engineering Group (EN-MME), CH-1211 Geneva 23 (Switzerland); Cerutti, F. [CERN, Engineering Department, Sources, Targets and Interactions Group (EN-STI), CH-1211 Geneva 23 (Switzerland); Charitonidis, N. [CERN, Engineering Department, Machines and Experimental Facilities Group (EN-MEF), CH-1211 Geneva 23 (Switzerland); École Polytechnique Fédérale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland); Charrondiere, C. [CERN, Engineering Department, Industrial Controls and Engineering Group (EN-ICE), CH-1211 Geneva 23 (Switzerland); Dallocchio, A.; Fernandez Carmona, P.; Francon, P.; Gentini, L.; Guinchard, M.; Mariani, N. [CERN, Engineering Department, Mechanical and Materials Engineering Group (EN-MME), CH-1211 Geneva 23 (Switzerland); Masi, A. [CERN, Engineering Department, Sources, Targets and Interactions Group (EN-STI), CH-1211 Geneva 23 (Switzerland); Marques dos Santos, S.D.; Moyret, P. [CERN, Engineering Department, Mechanical and Materials Engineering Group (EN-MME), CH-1211 Geneva 23 (Switzerland); Peroni, L. [Politecnico di Torino, Department of Mechanical and Aerospace Engineering (DIMEAS), Corso Duca degli Abruzzi 24, 10129 Torino (Italy); Redaelli, S. [CERN, Beams Department, Accelerators and Beams Physics Group (BE-ABP), CH-1211 Geneva 23 (Switzerland); Scapin, M. [Politecnico di Torino, Department of Mechanical and Aerospace Engineering (DIMEAS), Corso Duca degli Abruzzi 24, 10129 Torino (Italy)

    2013-08-01

    Predicting the consequences of highly energetic particle beams impacting protection devices as collimators or high power target stations is a fundamental issue in the design of state-of-the-art facilities for high-energy particle physics. These complex dynamic phenomena can be successfully simulated resorting to highly non-linear numerical tools (Hydrocodes). In order to produce accurate results, however, these codes require reliable material constitutive models that, at the extreme conditions induced by a destructive beam impact, are scarce and often inaccurate. In order to derive or validate such models a comprehensive, first-of-its-kind experiment has been recently carried out at CERN HiRadMat facility: performed tests entailed the controlled impact of intense and energetic proton pulses on a number of specimens made of six different materials. Experimental data were acquired relying on embedded instrumentation (strain gauges, temperature probes and vacuum sensors) and on remote-acquisition devices (laser Doppler vibrometer and high-speed camera). The method presented in this paper, combining experimental measurements with numerical simulations, may find applications to assess materials under very high strain rates and temperatures in domains well beyond particle physics (severe accidents in fusion and fission nuclear facilities, space debris impacts, fast and intense loadings on materials and structures etc.)

  5. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

    International Nuclear Information System (INIS)

    The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 418 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [1]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected 235U and 239Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as 236U capture. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for

  6. Universal Test Facility

    Science.gov (United States)

    Laughery, Mike

    A universal test facility (UTF) for Space Station Freedom is developed. In this context, universal means that the experimental rack design must be: automated, highly marketable, and able to perform diverse microgravity experiments according to NASA space station requirements. In order to fulfill these broad objectives, the facility's customers, and their respective requirements, are first defined. From these definitions, specific design goals and the scope of the first phase of this project are determined. An examination is first made into what types of research are most likely to make the UTF marketable. Based on our findings, the experiments for which the UTF would most likely be used included: protein crystal growth, hydroponics food growth, gas combustion, gallium arsenide crystal growth, microorganism development, and cell encapsulation. Therefore, the UTF is designed to fulfill all of the major requirements for the experiments listed above. The versatility of the design is achieved by taking advantage of the many overlapping requirements presented by these experiments.

  7. Advances in laser solenoid fusion reactor design

    International Nuclear Information System (INIS)

    The laser solenoid is an alternate fusion concept based on a laser-heated magnetically-confined plasma column. The reactor concept has evolved in several systems studies over the last five years. We describe recent advances in the plasma physics and technology of laser-plasma coupling. The technology advances include progress on first walls, inner magnet design, confinement module design, and reactor maintenance. We also describe a new generation of laser solenoid fusion and fusion-fission reactor designs

  8. Advances in Process Intensification through Multifunctional Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    O' Hern, Timothy [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Engineering Sciences Center; Evans, Lindsay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Materials Sciences and Engineering Center; Miller, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Materials Sciences and Engineering Center; Cooper, Marcia [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Energetic Components Realization Center; Torczynski, John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pena, Donovan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gill, Walt [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Engineering Sciences Center

    2011-02-01

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors utilizing pulse flow. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes associated with pulse flow for implementation in commercial applications. Sandia National Laboratories (SNL) operated a pilot-scale multifunctional reactor experiment for operation with and investigation of pulse flow operation. Validation-quality data sets of the fluid dynamics, heat and mass transfer, and chemical kinetics were acquired and shared with Chemical Research and Licensing (CR&L). Experiments in a two-phase air-water system examined the effects of bead diameter in the packing, and viscosity. Pressure signals were used to detect pulsing. Three-phase experiments used immiscible organic and aqueous liquids, and air or nitrogen as the gas phase. Hydrodynamic studies of flow regimes and holdup were performed for different types of packing, and mass transfer measurements were performed for a woven packing. These studies substantiated the improvements in mass transfer anticipated for pulse flow in multifunctional reactors for the acid-catalyzed C4 paraffin/olefin alkylation process. CR&L developed packings for this alkylation process, utilizing their alkylation process pilot facilities in Pasadena, TX. These packings were evaluated in the pilot-scale multifunctional reactor experiments established by Sandia to develop a more fundamental understanding of their role in process intensification. Lummus utilized the alkylation technology developed by CR&L to design and optimize the full commercial process utilizing multifunctional reactors containing the packings developed by CR&L and evaluated by Sandia. This hydrodynamic information has been developed for multifunctional chemical reactors utilizing pulse flow, for the acid-catalyzed C4 paraffin/olefin alkylation process, and is now accessible for use in

  9. High Energy Tests of Advanced Materials for Beam Intercepting Devices at CERN HiRadMat Facility

    CERN Document Server

    Bertarelli, A; Berthome, E; Boccone, V; Carra, F; Cerutti, F; Dallocchio, A; Dos Santos, S; Francon, P; Gentini, L; Guinchard, M; Mariani, N; Masi, A; Moyret, P; Redaeelli, S; Peroni, L; Scapin, M

    2012-01-01

    Predicting by simulations the consequences of LHC particle beams hitting Collimators and other Beam Intercepting Devices (BID) is a fundamental issue for machine protection: this can be done by resorting to highly non-linear numerical tools (Hydrocodes). In order to produce accurate results, these codes require reliable material models that, at the extreme conditions generated by a beam impact, are either imprecise or non-existent. To validate relevant constitutive models or, when unavailable, derive new ones, a comprehensive experimental test foreseeing intense particle beam impacts on six different materials, either already used for present BID or under development for future applications, is being prepared at CERN HiRadMat facility. Tests will be run at medium and high intensity using the SPS proton beam (440 GeV). Material characterization will be carried out mostly in real time relying on embarked instrumentation (strain gauges, microphones, temperature and pressure sensors) and on remote acquisition dev...

  10. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

    2008-02-01

    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  11. Claim prevention at reactor facilities

    International Nuclear Information System (INIS)

    Why does a radiation worker bring a claim alleging bodily injury from radiation exposure? Natural cancer, fear of radiation induced cancer, financial gain, emotional distress and mental anguish are some reasons for workers' claims. In this paper the author describes what power reactor health physicists are doing to reduce the likelihood of claims by establishing programs which provide sound protection of workers, prevent radiological events, improve workers' knowledge of radiological conditions and provide guidance for radiological incident response

  12. Effect of new cross-section evaluations on criticality and neutron energy spectrum of a typical material test research reactor

    International Nuclear Information System (INIS)

    Several new WIMSD libraries based on recent cross-section evaluations such as IAEA, ENDFB-VI, JENDL, and JEF have been made available by IAEA. These libraries were used for the computation of multiplication factor and energy spectrum for Pakistan Research Reactor-1 (PARR-1). Methodology was validated for benchmark problems made available by IAEA and comparison with reference results. The value of effective multiplication factors for all newly released libraries are 1.8-3.2% less than that of 1981 WIMSD library. The effect of various cross-section libraries on neutron energy spectrum was also studied. Differences of about -10% to 12.5% were found in thermal flux using the newly released libraries as compared with that obtained using 1981 WIMSD library. From the analysis, it was found that the main source of the difference is the cross-sections of hydrogen bound in water. When these cross-sections of hydrogen (bound in water) from new libraries were used along with all other data in 1981 WIMSD library, the keff obtained in this way has a difference of only 0.02-0.8% with that obtained from new libraries, while the flux spectrum agreed within 1% below 1 MeV with new libraries

  13. Effect of new cross-section evaluations on criticality and neutron energy spectrum of a typical material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, Siraj-ul-Islam [Pakistan Institute of Engineering and Applied Sciences (PIEAS), PO Nilore, Islamabad 45650 (Pakistan)]. E-mail: siraj1@hotpop.com; Ahmad, Nasir [Pakistan Institute of Engineering and Applied Sciences (PIEAS), PO Nilore, Islamabad 45650 (Pakistan); Aslam [Pakistan Institute of Engineering and Applied Sciences (PIEAS), PO Nilore, Islamabad 45650 (Pakistan)

    2004-11-01

    Several new WIMSD libraries based on recent cross-section evaluations such as IAEA, ENDFB-VI, JENDL, and JEF have been made available by IAEA. These libraries were used for the computation of multiplication factor and energy spectrum for Pakistan Research Reactor-1 (PARR-1). Methodology was validated for benchmark problems made available by IAEA and comparison with reference results. The value of effective multiplication factors for all newly released libraries are 1.8-3.2% less than that of 1981 WIMSD library. The effect of various cross-section libraries on neutron energy spectrum was also studied. Differences of about -10% to 12.5% were found in thermal flux using the newly released libraries as compared with that obtained using 1981 WIMSD library. From the analysis, it was found that the main source of the difference is the cross-sections of hydrogen bound in water. When these cross-sections of hydrogen (bound in water) from new libraries were used along with all other data in 1981 WIMSD library, the k{sub eff} obtained in this way has a difference of only 0.02-0.8% with that obtained from new libraries, while the flux spectrum agreed within 1% below 1 MeV with new libraries.

  14. CLIC Test Facility 3

    CERN Multimedia

    Kossyvakis, I; Faus-golfe, A

    2007-01-01

    The design of CLIC is based on a two-beam scheme, where short pulses of high power 30 GHz RF are extracted from a drive beam running parallel to the main beam. The 3rd generation CLIC Test Facility (CTF3) will demonstrate the generation of the drive beam with the appropriate time structure, the extraction of 30 GHz RF power from this beam, as well as acceleration of a probe beam with 30 GHz RF cavities. The project makes maximum use of existing equipment and infrastructure of the LPI complex, which became available after the closure of LEP.

  15. Preliminary design concept of an advanced integral reactor

    International Nuclear Information System (INIS)

    An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the reactor design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author)

  16. Hot helium flow test facility summary report

    International Nuclear Information System (INIS)

    This report summarizes the results of a study conducted to assess the feasibility and cost of modifying an existing circulator test facility (CTF) at General Atomic Company (GA). The CTF originally was built to test the Delmarva Power and Light Co. steam-driven circulator. This circulator, as modified, could provide a source of hot, pressurized helium for high-temperature gas-cooled reactor (HTGR) and gas-cooled fast breeder reactor (GCFR) component testing. To achieve this purpose, a high-temperature impeller would be installed on the existing machine. The projected range of tests which could be conducted for the project is also presented, along with corresponding cost considerations

  17. Reactor group constants and benchmark test

    International Nuclear Information System (INIS)

    The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

  18. Kaon: an advanced hadron facility

    International Nuclear Information System (INIS)

    An advanced hadron facility KAON has been proposed to be built in Canada. The report of the Project Definition Study has been presented to both levels of Government (federal and provincial) on May 24, 1990, for action in the near future. A short discussion will be given of the scientific motivation. The physics along the intensity and precision frontier is fully complementary to the physics along the energy frontier. Following, a description will be given of the 100 μA, 30 GeV proton synchrotron proposed. The accelerator will consist of five rings using the present 500 MeV cyclotron as an injector. If the project were funded this year, the accelerators would be completed by 1995 or so, with the experimental program starting a year later

  19. ETRR-2 nuclear reactor: Facility specification

    International Nuclear Information System (INIS)

    The report provides technical details on the ETRR-2 nuclear reactor core and immediate structure for analysis purposes. The goal of the report is to provide sufficient geometric and material data to build a computational neutronic model of the facility. (author)

  20. Advanced nuclear reactor public opinion project

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  1. Advanced nuclear reactor public opinion project

    International Nuclear Information System (INIS)

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions

  2. Low-temperature irradiation facilities in reactors

    International Nuclear Information System (INIS)

    The general principle of low-temperature irradiation facilities in reactors is described and the main fields of research which are relevant to them are discussed. Existing facilities using He, H2, Ne or N2 as cooling liquids (or gases) are compared and some special problems as the production of nuclear heat, the choice of the cooling agent, safety considerations and continuous local flux monitors are spelled out. Finally, the liquid He facility in Garching is described along with some of the experimental techniques used and experiments performed. (K.B.)

  3. Experimental investigation of thermal limits in parallel plate configuration for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    The Advanced Neutron Source Reactor (ANSR) is currently being designed to become the world's highest-flux, steady-state, thermal neutron source for scientific experiments. Highly subcooled, heavy-water coolant flows vertically upward at a very high velocity of 25 m/s through parallel aluminum fuel-plates. The core has average and peak heat fluxes of 5.9 and 12 MW/m2, respectively. In this configuration, both flow excursion (FE) and true critical heat flux (CHF), represent potential thermal limitations. The availability of experimental data for both FE and true CHF at the conditions applicable to the ANSR is very limited. A Thermal Hydraulic Test Loop (THTL) facility was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of both thermal limits under the expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 14 MW/m2 and a corresponding velocity range of 8 to 21 m/s. Both the exit pressure (1.7 MPa) and inlet temperature (45 degrees C) were maintained constant for these tests, while the loop was operated in a ''stiff''(constant flow) mode. Limited experiments were also conducted at 12 MW/m2 using a ''soft'' mode (near constant pressure-drop) for actual FE burnout tests and using a ''stiff' mode for true CHF tests, to compare with the original FE experiments

  4. Fast critical experiment data for space reactors

    International Nuclear Information System (INIS)

    Data from a number of previous critical experiments exist that are relevant to the design concepts being considered for SP-100 and MMW space reactors. Although substantial improvements in experiment techniques have since made some of the measured quantities somewhat suspect, the basic criticality data are still useful in most cases. However, the old experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, we have calculated about 20 fast benchmark critical experiments with the latest ENDF/B data and modern transport codes. These calculations were undertaken as a part of the planning process for a new series of benchmark experiments aimed at supporting preliminary designs of SP-100 and MMW space reactors

  5. Scaling philosophy and system description of AHWR Thermal-Hydraulic Test Facility (ATTF)

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) being designed in India is a 920 MWth pressure tube type boiling light water cooled and heavy water moderated reactor. AHWR Thermal Hydraulic Test Facility (ATTF), a scaled experimental facility that simulates the thermal-hydraulic behaviour of main heat transport system and ECCS, is designed. The objectives of the facility are to obtain thermal margin (CHF) and the parallel channel stability behaviour Global scaling is based on Power to Volume ratio. This philosophy is based on maintaining the same pressure, temperature with same working fluid. Main advantage of this scaling approach is that it preserves the time scales which are very crucial for the simulation of transient and accident conditions to assess the performance of safety systems. All of the Main Heat Transport (MHT) and Emergency Core Cooling System (ECCS) components are scaled down on the basis of power to volume scaling. ATTF contains two full power channels in comparison with 452 channels of AHWR then the scaling ratio is 226. Therefore the volumes of the components in natural circulation path (MHT) are scaled down by 226. Different local phenomenon like Critical Heat Flux (CHF), Flashing, Geysering etc which affects the performance of the system are scaled down appropriately. GDCS injection, feed water flow etc are simulated as boundary flow scaling approach. This 3-level approach simulates almost all the thermal hydraulics phenomenon of the prototype in the model, with the appropriate scale of the model to the prototype. (author)

  6. Jordan Research and Training Reactor (JRTR) Utilization Facilities

    International Nuclear Information System (INIS)

    Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U3Si2) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×1014 cm-2s-1, and is controlled by a Hafnium control absorber rod and B4C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

  7. Thermochemistry of nuclear fuels in advanced reactors

    International Nuclear Information System (INIS)

    The presence of a large number of elements, accompanied with steep temperature gradient results in dynamic chemistry during nuclear fuel burn-up. Understanding this chemistry is very important for efficient and safe usage of nuclear fuels. The radioactive nature of these fuels puts lot of constraint on regulatory bodies to ensure their accident free operation in the reactors. One of the common aims of advanced fuels is to achieve high burn-up. As burn-up of the fuel increases, chemistry of fission-products becomes increasingly more important. To understand different phenomenon taking place in-pile, many out of-pile experiments are carried out. Extensive studies of thermodynamic properties, phase analysis, thermophysical property evaluation, fuel-fission product clad compatibility are carried out with relevant compounds and simulated fuels (SIMFUEL). All these data are compiled and jointly evaluated using different computational methods to predict fuel behaviour during burn-up. Only when this combined experimental and theoretical information confirms safe operation of the pin, a test pin is prepared and burnt in a test reactor. Every fuel has a different chemistry and different constraints associated with it. In this talk, various thermo-chemical aspects of some of the advanced fuels, mixed carbide, mixed nitride, 'Pu' rich MOX, 'Th' based AHWR fuels and metallic fuels will be discussed. (author)

  8. Test facility TIMO for testing the ITER model cryopump

    International Nuclear Information System (INIS)

    Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L' Air Liquide in the form of a NET contract. (author)

  9. Applications of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    When the technique of neutron scattering was pioneered at the X-10 graphite reactor at Oak Ridge National Laboratory about 50 years ago, it was used to study certain important, but fairly esoteric, properties of crystals. From this modest beginning, neutron scattering has become a major tool in every branch of science, from the astrophysics of the early universe to human biology, and in many important industrial and engineering applications. In a typical modern research reactor it is not unusual to find one instrument studying new polymeric materials, while its neighbor is measuring residual stress in a jet turbine, sometimes with the jet operating. Most of this development has taken place outside of the United States, primarily in Western Europe, Japan and Russia, and it is generally recognized that we are a decade behind our competitors in this important field. The Advanced Neutron Source (ANS), planned to become operational as a user-facility at Oak Ridge at the end of this decade, will regain our leadership in neutron-based research and will be a major center for attracting new students into science. This paper discusses some of the research and development applications of the ANS, with an emphasis on applied materials science and engineering

  10. Utilization of the BARC critical facility for ADS related experiments

    Indian Academy of Sciences (India)

    Rajeev Kumar; R Srivenkatesan

    2007-02-01

    The paper discusses the basic design of the critical facility, whose main purpose is the physics validation of AHWR. Apart from moderator level control, the facility will have shutdown systems based on shutoff rods and multiple ranges of neutron detection systems. In addition, it will have a flux mapping system based on 25 fission chambers, distributed in the core. We are planning to use this reactor for experiments with a suitable source to simulate an ADS system. Any desired sub-criticality can be achieved by adjusting the moderator level. Apart from perfecting our experimental techniques, in simple configurations, we intend to study the one-way coupled core in this facility. Preliminary calculations, employing a Monte Carlo code TRIPOLI, are presented.

  11. Advances in reactor safety research

    International Nuclear Information System (INIS)

    The Nuclear Safety Project is an important part of the German reactor safety research programme. It works on problems concerning safety and environemental risks of LWR reactors and reprocessing plants and investigates accident consequences. At the 1978 annual meeting, the core behaviour on cooling and reactivity disturbances was discussed, as well as release, retention, and possible radiological effects of radioactive pollutants. Among other subjects, fission product retention in LWR reactors and reprocessing plants were reported on as well as hypothetic core meltdown. (orig.)

  12. Status of advanced technologies for CANDU reactors

    International Nuclear Information System (INIS)

    The future development of the CANDU reactor is a continuation of a successful series of reactors, the most recent of which are nine CANDU 6 Mk 1* units and four Darlington units. There are three projects underway that continue the development of the CANDU reactor. These new design projects flow from the original reactor designs and are a natural progression of the CANDU 6 Mk 1, two units of which are operating successfully in Canada, one each in Argentina and Korea, with five more being built in Rumania. These new design projects are known as: CANDU 6 Mk 2, an improved version of CANDU 6 Mk 1; CANDU 3, a small, advanced version of the CANDU 6 Mk 1; CANDU 6 Mk 3, a series of advanced CANDU reactors. A short description of modified versions of CANDU reactors is given in this paper. 5 figs

  13. Post reactor researches of fuel pins, tested under alternating NEMF reactor functioning modes

    International Nuclear Information System (INIS)

    Changing of rod ceramic fuel pins state under their exploitation conditions changing influence at alternating of three-mode nuclear energy-moving facility reactor functioning has been examined. There are presented the results of researches of fuel pins, tested in the reactor IRGIT and RA, firstly under moving mode, then - under energy mode of minor power of NEMF reactor. (author)

  14. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report December 2014

    Energy Technology Data Exchange (ETDEWEB)

    Renae Soelberg

    2014-12-01

    • PNNL has completed sectioning of the U.C. Berkeley hydride fuel rodlet 1 (highest burn-up) and is currently polishing samples in preparation for optical metallography. • A disk was successfully sectioned from rodlet 1 at the location of the internal thermocouple tip as desired. The transition from annular pellet to solid pellet is verified by the eutectic-filled inner cavity located on the back face of this disk (top left) and the solid front face (bottom left). Preliminary low-resolution images indicate interesting sample characteristics in the eutectic surrounding the rodlet at the location of the outer thermocouple tip (right). This sample has been potted and is currently being polished for high-resolution optical microscopy and subsequent SEM analysis. (See images.)

  15. TRIGA MARK II first research reactor facility in Kingdom of Morocco

    International Nuclear Information System (INIS)

    The research reactor facility is located at Centre d'Etudes Nucleaires de la Maamora(CENM), located approximately 25 kilometers north of the city of Rabat. This facility will enable CNESTEN, as the operating organization, to fulfil its missions for promotion of nuclear technology in Morocco, contribute to the implementation of a national nuclear power program, and assist the state in monitoring nuclear activities for protection of the public and environment. The reactor building include TRIGA Mark II research reactor with an initial power level of 2000kW (t), and equipped for a planned future upgrade to 3,000-kilowatts.The facility is the keystone structure of CENM, and contain in addition to the TRIGA research reactor, extensively equipped laboratories and all associate support systems, structures, and supply facilities with the support of the AIEA, French CEA and LLNL (USA). The CENM with its TRIGA reactor and fully equipped laboratories will give the kingdom of Morocco its first nuclear installation with extensive capabilities. These will include the production of radioisotopes for medical, industrial and environmental uses, metallurgy and chemistry, implementation of nuclear analytical techniques such as neutron activation analysis and non-destructive examination techniques, as well as carrying out basic research programs in solid state and reactor physics. The feedback from the commissioning and the implementation of the safety standards during this phase was very interesting from safety point of view. The TRIGA Mark II research reactor at CENM achieved initial criticality on May 2, 2007 at 13:30 with 71 fuel elements and culminated with the successful completion of the full power endurance testing on 6 September, 2007.

  16. Fast Breeder Test Reactor: 15 years of operating experience

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 1985 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 1993 and power was raised to 10.5 MWt in Dec 1993. Turbine generator was synchronized to the grid in Jul 1997. The indigenously developed mixed carbide fuel has achieved a peak burn up of 88,000 MWd/t till now at a linear heat rating of 320 W/cm and reactor power of 13.4 MWt without any fuel-clad failure. The paper presents operating and decontamination experience, performance of fuel, steam generator and sodium circuits, certain unusual occurrences encountered by the plant and various improvements carried out in reactor systems to enhance plant availability. (author)

  17. Radiation exposure doses of employees in reactor facilities for test and research and under research and development stages, and in facilities for nuclear fuel refining, fabrication, reprocessing and usage

    International Nuclear Information System (INIS)

    (1) Radiation exposure doses in reactor facilities. The owners of reactor facilities are obliged by law to control the radiation exposure doses of the employees below the permissible levels. The data based on the reports made in this connection are given in tables for the fiscal year 1978 (from April 1978 to March 1979). It was revealed that the radiation exposure doses of the employees were far below the permissible levels. The distributions of exposure doses in Japan Atomic Energy Research Institute, Power Reactor and Nuclear Fuel Development Corporation and so on are presented for the whole year and the respective quarters. (2) Radiation exposure doses in facilities for nuclear fuel. The owners are similarly obliged to control radiation exposure. The data in this connection are given, and the doses were far below the permissible levels. The distributions in the private enterprises and so on are presented for the whole year. (J.P.N.)

  18. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  19. Preliminary design concepts of an advanced integral reactor

    International Nuclear Information System (INIS)

    An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

  20. Air gun test facility

    International Nuclear Information System (INIS)

    This paper describes a facility that is potentially useful in providing data for models to predict the effects of nuclear explosions on cities. IIT Research Institute has a large air gun facility capable of launching heavy items of a wide variety of geometries to velocities ranging from about 80 fps to 1100 fps. The facility and its capabilities are described, and city model problem areas capable of investigation using the air gun are presented

  1. Analytical chemistry requirements for advanced reactors

    International Nuclear Information System (INIS)

    The nuclear power industry has been developing and improving reactor technology for more than five decades. Newer advanced reactors now being built have simpler designs which reduce capital cost. The greatest departure from most designs now in operation is that many incorporate passive or inherent safety features which require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. India is developing the Advanced Heavy Water Reactor (AHWR) in its plan to utilise thorium in nuclear power program

  2. ANDREA: Advanced nodal diffusion code for reactor analysis

    International Nuclear Information System (INIS)

    A new macro code is being developed at NRI which will allow coupling of the advanced thermal-hydraulics model with neutronics calculations as well as efficient use in core loading pattern optimization process. This paper describes the current stage of the macro code development. The core simulator is based on the nodal expansion method, Helios lattice code is used for few group libraries preparation. Standard features such as pin wise power reconstruction and feedback iterations on critical control rod position, boron concentration and reactor power are implemented. A special attention is paid to the system and code modularity in order to enable flexible and easy implementation of new features in future. Precision of the methods used in the macro code has been verified on available benchmarks. Testing against Temelin PWR operational data is under way (Authors)

  3. Startup testing of Romania dual-core test reactor

    International Nuclear Information System (INIS)

    Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

  4. Review of the Advanced Neutron Source (ANS) materials irradiation facilities

    International Nuclear Information System (INIS)

    The purpose of the workshop was to document as accurately as possible the present and future needs for neutron irradiation capacity and facilities as related to the design of the Advanced Neutron Source (ANS) which will be the next generation steady-state research reactor. The report provides the findings and recommendations of the working group. After introductory and background information is presented, the discussion includes the status of the ANS design, in particular in-core materials irradiation facilities design and important experimental parameters. The summary of workshop discussions describes a survey of irradiation-effects research community and opportunities for ex-core irradiation facilities. 20 refs., 2 figs., 4 tabs

  5. Systematics of Reconstructed Process Facility Criticality Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Pruvost, N.L.; McLaughlin, T.P.; Monahan, S.P.

    1999-09-19

    The systematics of the characteristics of twenty-one criticality accidents occurring in nuclear processing facilities of the Russian Federation, the United States, and the United Kingdom are examined. By systematics the authors mean the degree of consistency or agreement between the factual parameters reported for the accidents and the experimentally known conditions for criticality. The twenty-one reported process criticality accidents are not sufficiently well described to justify attempting detailed neutronic modeling. However, results of classic hand calculations confirm the credibility of the reported accident conditions.

  6. Decommissioning of reactor facilities (2). Required technology

    International Nuclear Information System (INIS)

    Decommissioning of reactor facilities was planned to perform progressive dismantling, decontamination and radioactive waste disposal with combination of required technology in a safe and economic way. This article outlined required technology for decommissioning as follows: (1) evaluation of kinds and amounts of residual radioactivity of reactor facilities with calculation and measurement, (2) decontamination technology of metal components and concrete structures so as to reduce worker's exposure and production of radioactive wastes during dismantling, (3) dismantling technology of metal components and concrete structures such as plasma arc cutting, band saw cutting and controlled demolition with mostly remote control operation, (3) radioactive waste disposal for volume reduction and reuse, and (4) project management of decommissioning for safe and rational work to secure reduction of worker's exposure and prevent the spreading of contamination. (T. Tanaka)

  7. Heavy Water Components Test Reactor Decommissioning

    International Nuclear Information System (INIS)

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D and D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  8. A facile surfactant critical micelle concentration determination

    OpenAIRE

    Cai, Lifeng; Gochin, Miriam; Liu, Keliang

    2011-01-01

    Liquid surface curvature variations in microplate wells due to different liquid surface tension cause significant signal change in spectroscopic measurement using a plate reader with a vertical detecting light beam. The signals have been quantitated and used to develop a method for facile surfactant critical micelle concentration determination.

  9. The integral test facility Karlstein - INKA

    International Nuclear Information System (INIS)

    The INKA (INtegral Test Facility KArlstein) test facility was designed and erected to test and demonstrate performance of the passive safety systems of KERENA™, the new AREVA Boiling Water Reactor (BWR) design. The experimental program within the KERENA™ development program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests will be performed to simulate transients and LOCA (Loss of Coolant Accident) scenarios at the INKA test facility. These tests will test and demonstrate the interaction between the passive components/systems and demonstrate their ability to perform their design function. For the integral tests, the Passive Pressure Pulse Transmitter will be included. The INKA test facility represents the KERENA™ Containment with a volume scaling of 1:24. Component heights and levels are full scale in order to match the driving forces for natural circulation. The reactor pressure vessel is simulated by the accumulator vessel of the large valve test facility of Karlstein - a vessel with a design pressure of 11 MPa and a storage capacity of 125 m3. The vessel is fed by a benson boiler with a maximum power supply of 22 MW. The drywell of the INKA containment is divided into two compartments and connected to the wetwell (Pressure Suppression System) via a full scale vent pipe. Therefore, the INKA pressure suppression system meets the requirements of modern and existing BWR designs. As a result of the large power supply at the facility, INKA is capable of simulating various accident scenarios starting with the initiating event - for example pipe rupture. At INKA a full train of passive safety systems is available. INKA is also able to simulate the functions of active safety system such as containment heat removal. Therefore accident scenarios relevant to modern Gen III as well as for operating Gen II design can be simulated in order to validate system and

  10. Typical technology of mechanics on Gen-III passive NPPs and Gen-IV advanced supercritical light water reactors

    International Nuclear Information System (INIS)

    Full text: Technical requirements for Gen-III advanced nuclear power plants, which take passive reactors as the main body, were originally brought forward in American 'Advanced Light Water Reactor Utility Requirement Document' (ALWR-URD) in early 1990's. The primary characteristic of passive nuclear power plant is large amount of simplification to the original active safety systems, replacing or supplementing them with passive safety systems, which enhances safety and economy. However, the replacement of active safety systems by passive safety systems also brings about some mechanics that compel attention, typically, such as load-carrying capability evaluation for steel containment, in-vessel retention (IVR) of molten core debris, seismic design without OBE, thermo-hydraulic issues concerning with coupling between two-phase fluid and solid, etc. At the beginning of this century, six typical Gen-IV advanced reactor types (Sodium Cooled Fast Reactor, Supercritical Water-Cooled Reactor, etc.) were put forward. Among these types of reactors, Supercritical Water-Cooled Reactor adopts supercritical water as coolant and operates above the thermodynamic critical point of water by increasing temperature and pressure of the coolant, which makes the plant economic and efficient. However, this type of reactor also brings about some mechanical difficulties (e.g. pressure fluctuation caused by the supercritical fluid in the core, creep of materials working at high temperature, etc.) for the design of facility and components. In this paper, the issues mentioned above are outlined for further consideration. (author)

  11. BFS, a Legacy to the International Reactor Physics, Criticality Safety, and Nuclear Data Communities

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Anatoly Tsibulya; Yevgeniy Rozhikhin

    2012-03-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. Data provided by these two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades The Russian Federation has been a major contributor to both projects with the Institute of Physics and Power Engineering (IPPE) as the major contributor from the Russian Federation. Included in the benchmark specifications from the BFS facilities are 34 critical configurations from BFS-49, 61, 62, 73, 79, 81, 97, 99, and 101; spectral characteristics measurements from BFS-31, 42, 57, 59, 61, 62, 73, 97, 99, and 101; reactivity effects measurements from BFS-62-3A; reactivity coefficients and kinetics measurements from BFS-73; and reaction rate measurements from BFS-42, 61, 62, 73, 97, 99, and 101.

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

    Energy Technology Data Exchange (ETDEWEB)

    Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

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

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

    International Nuclear Information System (INIS)

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

  14. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  15. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

  16. Small Satellite Space Environments Effects Test Facility

    OpenAIRE

    Dennison, JR; Hartley, Kent; Montierth Phillipps, Lisa; Johnson, Robert; Dekany, Justin; Dyer, James

    2014-01-01

    A versatile space environments test facility has been designed and built to study the effects on small satellites and system components. Testing for potentially environmental-induced modifications of small satellites is critical to avoid possible deleterious or catastrophic effects over the duration of space missions. This is increasingly more important as small satellite programs have longer mission lifetimes, expand to more hash env...

  17. Advanced toroidal facility vaccuum vessel stress analyses

    International Nuclear Information System (INIS)

    The complex geometry of the Advance Toroidal Facility (ATF) vacuum vessel required special analysis techniques in investigating the structural behavior of the design. The response of a large-scale finite element model was found for transportation and operational loading. Several computer codes and systems, including the National Magnetic Fusion Energy Computer Center Cray machines, were implemented in accomplishing these analyses. The work combined complex methods that taxed the limits of both the codes and the computer systems involved. Using MSC/NASTRAN cyclic-symmetry solutions permitted using only 1/12 of the vessel geometry to mathematically analyze the entire vessel. This allowed the greater detail and accuracy demanded by the complex geometry of the vessel. Critical buckling-pressure analyses were performed with the same model. The development, results, and problems encountered in performing these analyses are described. 5 refs., 3 figs

  18. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  19. Research reactor and fuel development/production facility decommissioning technology and experience

    International Nuclear Information System (INIS)

    This paper discusses the technology and experience gained in a series of reactor and fuels development facility decommissioning programs carried out by Babcock and Wilcox (B and W) at its US Nuclear Regulatory Commission (NRC)-licensed sites in Lynchburg, Virginia. Areas of generic application to future projects are particularly emphasized. The projects included one test and one research reactor, four low-power critical experiment facilities, and two buildings that housed plutonium/uranium fuels development laboratories. These projects were comprehensive; they included developing the decommissioning and quality assurance plans, interfacing with the NRC, performing the actual decontamination/dismantling work, performing predecontamination and final radiological surveys, and volume reducing, packaging, certifying, classifying and shipping the radioactive waste for disposal

  20. Material test reactor fuel research at the BR2 reactor

    International Nuclear Information System (INIS)

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor of