Sample records for reactor design certification

  1. Structural Design Challenges in Design Certification Applications for New Reactors

    SciTech Connect (OSTI)

    Miranda, M.; Braverman, J.; Wei, X.; Hofmayer, C.; Xu, J.

    2011-07-17T23:59:59.000Z

    The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are confined within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of structural design chal- lenges encountered in recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

  2. Seismic Analysis Issues in Design Certification Applications for New Reactors

    SciTech Connect (OSTI)

    Miranda, M.; Morante, R.; Xu, J.

    2011-07-17T23:59:59.000Z

    The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of several seismic analysis issues encountered during a review of recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

  3. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    SciTech Connect (OSTI)

    Not Available

    1994-07-01T23:59:59.000Z

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  4. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    SciTech Connect (OSTI)

    Not Available

    1994-07-01T23:59:59.000Z

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  5. Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

    SciTech Connect (OSTI)

    NONE

    1997-05-01T23:59:59.000Z

    This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff`s review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design.

  6. Advanced light water reactor plants system 80+{trademark} design certification program. Annual progress report, October 1, 1993--September 30, 1994

    SciTech Connect (OSTI)

    Not Available

    1995-01-01T23:59:59.000Z

    The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80{sup +}{trademark} during the U.S. government`s 1994 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2 and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems. Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units and the System 80+ design form the basis of the Korean standardization program. The Nuclear Island portion of the System 80+ standard design has also been offered to the Republic of China, in response to their bid specification for an ALWR. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was docketed by the Nuclear Regulatory Commission (NRC) in May 1991 and a Draft Safety Evaluation Report (DSER) was issued in October 1992.

  7. Certification plan for reactor analysis computer codes

    SciTech Connect (OSTI)

    Toffer, H.; Crowe, R.D.; Schwinkendorf, K.N. [Westinghouse Hanford Co., Richland, WA (United States); Pevey, R.E. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1990-01-01T23:59:59.000Z

    A certification plan for reactor analysis computer codes used in Technical Specifications development and for other safety and production support calculations has been prepared. An action matrix, checklists, a time schedule, and a resource commitment table have been included in the plan. These items identify what is required to achieve certification of the codes, the time table that this will be accomplished on, and the resources needed to support such an effort.

  8. System 80+{trademark} Standard Design: CESSAR design certification. Volume 4: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This report, Volume 4, provides a description of the reactor, reactor internals, fuel assemblies, and associated design requirements.

  9. Geology Department Graduate Certificates: These certificates are designed to provide practicing professionals an opportunity to

    E-Print Network [OSTI]

    Geology Department Graduate Certificates: These certificates are designed to provide practicing are offered: Certificate in Engineering Geology Purpose The Graduate Certificate in Engineering Geology provides practicing geologists an opportunity to upgrade their engineering geology credentials while

  10. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-08-01T23:59:59.000Z

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  11. ap600 design certification: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    for the future The Green Building and Sustainable Design Certificate Program addresses Energy Storage, Conversion and Utilization Websites Summary: Building for the future The...

  12. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-06-01T23:59:59.000Z

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

  13. Designing Reactors to Facilitate Decommissioning

    SciTech Connect (OSTI)

    Richard H. Meservey

    2006-06-01T23:59:59.000Z

    Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

  14. Designing decommissioning into new reactor designs

    SciTech Connect (OSTI)

    Devgun, J.S.; CHMM, Ph.D. [Nuclear Power Technologies, Sargent and Lundy LLC, Chicago, IL (United States)

    2007-07-01T23:59:59.000Z

    One of the lessons learned from decommissioning of existing reactors has been that decommissioning was not given much thought when these reactors were designed some three or four decades ago. Recently, the nuclear power has seen a worldwide resurgence and many new advanced reactor designs are either on the market or nearing design completion. Most of these designs are evolutionary in nature and build on the existing and proven technologies. They also incorporate many improvements and take advantage of the substantial operating experience. Nevertheless, by and large, the main factors driving the design of new reactors are the safety features, safeguards considerations, and the economic factors. With a large decommissioning experience that already exists in the nuclear industry, and with average decommissioning costs at around six hundred million dollars for each reactor in today's dollars, it is necessary that decommissioning factors also be considered as a part of the early design effort. Even though decommissioning may be sixty years down the road from the time they go on line, it is only prudent that new designs be optimized for eventual decommissioning, along with the other major considerations. (authors)

  15. Advanced burner test reactor preconceptual design report.

    SciTech Connect (OSTI)

    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

    2008-12-16T23:59:59.000Z

    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 advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

  16. Nuclear Reactor Safety Design Criteria

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1993-01-19T23:59:59.000Z

    The order establishes nuclear safety criteria applicable to the design, fabrication, construction, testing, and performance requirements of nuclear reactor facilities and safety class structures, systems, and components (SSCs) within these facilities. Cancels paragraphs 8a and 8b of DOE 5480.6. Cancels DOE O 5480.6 in part. Certified 11-18-10.

  17. Pollution prevention through reactor design

    SciTech Connect (OSTI)

    Hopper, J.R. [Lamar Univ., Beaumont, TX (United States)

    1995-09-01T23:59:59.000Z

    Generation of waste in the chemical processing industries has its beginning in the heart of the process--the reaction system. Pollution prevention will have the greatest impact in minimizing the generation of waste through the design and operation of chemical reactors by reducing generation at the source--source reduction. Pollution prevention by modification of reaction parameters is defined as changing the selectivity of the reaction so that undesirable reactions which produce waste products are minimized while at the same time producing the desirable products.

  18. System 80+{trademark} Standard Design: CESSAR design certification. Volume 2: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This report, Volume 2, in conjunction with Volume 3, provides the design of structures, components, equipment and systems.

  19. System 80+{trademark} Standard Design: CESSAR design certification. Volume 3: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This report, Volume 3, in conjunction with Volume 2, provides the design of structures, components, equipment and systems.

  20. Construction Certifications

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Certifications Individual Permit: Construction Certifications Enhanced control measures were designed and installed at 67 Sites in 42 SMAs in 2012. Contact Environmental...

  1. Reactor protection system design alternatives for sodium fast reactors

    E-Print Network [OSTI]

    DeWitte, Jacob D. (Jacob Dominic)

    2011-01-01T23:59:59.000Z

    Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a ...

  2. Reactor physics design of supercritical CO?-cooled fast reactors

    E-Print Network [OSTI]

    Pope, Michael A. (Michael Alexander)

    2004-01-01T23:59:59.000Z

    Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO? (S-CO?) as a Brayton cycle working fluid in a direct ...

  3. System 80+{trademark} Standard Design: CESSAR design certification. Volume 6: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This report, Volume 6, in conjunction with Volume 7, provides a description of engineered safety features.

  4. System 80+{trademark} Standard Design: CESSAR design certification. Volume 11: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 11 discusses Radiation Protection, Conduct of Operations, and the Initial Test Program.

  5. System 80+{trademark} Standard Design: CESSAR design certification. Volume 8: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 8 provides a description of instrumentation and controls.

  6. System 80+{trademark} Standard Design: CESSAR design certification. Volume 12: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 12, along with volume 13, provides accident analyses.

  7. System 80+{trademark} Standard Design: CESSAR design certification. Volume 1: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These Volumes, describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This document, Volume 1, provides an introduction and general description of plant and site envelope characteristics.

  8. System 80+{trademark} Standard Design: CESSAR design certification. Volume 17: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describes the Combustion Engineering, Inc. System 80+{trademark} Standard Design. This Volume 17 provides Appendix A of this report, closure of unresolved and Genetic Safety Issues.

  9. System 80+{trademark} Standard Design: CESSAR design certification. Volume 18: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80+{trademark} Standard Design. This Volume 18 provides Appendix B, Probabilistic Risk Assessment.

  10. System 80+{trademark} Standard Design: CESSAR design certification. Volume 9: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 9 discusses Electric Power and Auxiliary Systems.

  11. System 80+{trademark} Standard Design: CESSAR design certification. Volume 10: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 10 discusses the Steam and Power Conversion System and Radioactive Waste Management.

  12. Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1994-12-31T23:59:59.000Z

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials.

  13. Mirror Advanced Reactor Study interim design report

    SciTech Connect (OSTI)

    Not Available

    1983-04-01T23:59:59.000Z

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

  14. Generic small modular reactor plant design.

    SciTech Connect (OSTI)

    Lewis, Tom Goslee,; Cipiti, Benjamin B.; Jordan, Sabina Erteza; Baum, Gregory A.

    2012-12-01T23:59:59.000Z

    This report gives an overview of expected design characteristics, concepts, and procedures for small modular reactors. The purpose of this report is to provide those who are interested in reducing the cost and improving the safety of advanced nuclear power plants with a generic design that possesses enough detail in a non-sensitive manner to give merit to their conclusions. The report is focused on light water reactor technology, but does add details on what could be different in a more advanced design (see Appendix). Numerous reactor and facility concepts were used for inspiration (documented in the bibliography). The final design described here is conceptual and does not reflect any proposed concept or sub-systems, thus any details given here are only relevant within this report. This report does not include any design or engineering calculations.

  15. Design options for a bunsen reactor.

    SciTech Connect (OSTI)

    Moore, Robert Charles

    2013-10-01T23:59:59.000Z

    This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

  16. Spheromak reactor-design study

    SciTech Connect (OSTI)

    Les, J.M.

    1981-06-30T23:59:59.000Z

    A general overview of spheromak reactor characteristics, such as MHD stability, start up, and plasma geometry is presented. In addition, comparisons are made between spheromaks, tokamaks and field reversed mirrors. The computer code Sphero is also discussed. Sphero is a zero dimensional time independent transport code that uses particle confinement times and profile parameters as input since they are not known with certainty at the present time. More specifically, Sphero numerically solves a given set of transport equations whose solutions include such variables as fuel ion (deuterium and tritium) density, electron density, alpha particle density and ion, electron temperatures.

  17. Building for the future The Green Building and Sustainable Design Certificate Program addresses

    E-Print Network [OSTI]

    California at Davis, University of

    Building for the future The Green Building and Sustainable Design Certificate Program addresses for implementation of green building and sustainable design into site planning and design, building design and construction, and building management practices. Started as a joint program between two of UC Davis Extension

  18. Status of fusion reactor blanket design

    SciTech Connect (OSTI)

    Smith, D.L.; Sze, D.K.

    1986-11-01T23:59:59.000Z

    This paper provides a brief review of the Blanket Comparison and Selection Study (BCSS)/sup 1/ and an overview of more recent fusion reactor blanket design efforts. Specific areas covered include improvements in leading blanket concepts identified in the BCSS, viz., self-cooled liquid metal concepts, helium-cooled solid breeder concepts, and helium-cooled liquid breeder concepts. In addition, a summary of innovative blanket concepts and design features is presented. The key features and critical issues associated with these designs are identified.

  19. Initial Design of a Dual Fluidized Bed Reactor

    E-Print Network [OSTI]

    Yun, Minyoung

    2014-01-01T23:59:59.000Z

    bed, Chemical Engineering Research and Design, Volume 90,Design of a Dual Fluidized Bed Reactor by Minyoung Yun Master of Science, Graduate Program in Chemical and Environmental Engineering

  20. Innovative design of uranium startup fast reactors

    E-Print Network [OSTI]

    Fei, Tingzhou

    2012-01-01T23:59:59.000Z

    Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

  1. Secretary Chu Statement on AP1000 Reactor Design Certification | Department

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the.pdfBreakingMay 2015ParentsMiddle SchoolPhysics | DepartmentRecovery ActofGEEnergyof

  2. Secretary Chu Statement on AP1000 Reactor Design Certification | Department

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOriginEducationVideo »Usage »DownloadSolarSequestrationof the Americas |ResponseSpeaksof

  3. Final safety evaluation report related to the certification of the System 80{sup +} design (Docket No. 52-002). Volume 2, Chapters 15--22 and appendices

    SciTech Connect (OSTI)

    Not Available

    1994-08-01T23:59:59.000Z

    This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the system 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of ABB-CE`s System 80 design from which it evolved. Unique features of the System 80+ design include: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 2, contains Chapters 15 through 22 and Appendices A through E.

  4. Final safety evaluation report related to the certification of the System 80{sup +} design (Docket No. 52-002). Volume 1, Chapters 1--14

    SciTech Connect (OSTI)

    Not Available

    1994-08-01T23:59:59.000Z

    This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the System 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of Abb-CE`s System 80 design from which it evolved. Unique features of the System 80+ design included: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors, and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 1, contains Chapters 1 through 14 of this report.

  5. NRC policy on future reactor designs

    SciTech Connect (OSTI)

    none,

    1985-07-01T23:59:59.000Z

    On April 13, 1983, the US Nuclear Regulatory Commission issued for public comment a ''Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation'' (48 FR 16014). This report presents and discusses the Commission's final version of that policy statement now entitled, ''Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants.'' It provides an overview of comments received from the public and the Advisory Committee on Reactor Safeguards and the staff response to these. In addition to the Policy Statement, the report discusses how the policies of this statement relate to other NRC programs including the Severe Accident Research Program; the implementation of safety measures resulting from lessons learned in the accident at Three Mile Island; safety goal development; the resolution of Unresolved Safety Issues and other Generic Safety Issues; and possible revisions of rules or regulatory requirements resulting from the Severe Accident Source Term Program. Also discussed are the main features of a generic decision strategy for resolving Regulatory Questions and Technical Issues relating to severe accidents; the development and regulatory use of new safety information; the treatment of uncertainty in severe accident decision making; and the development and implementation of a Systems Reliability Program for both existing and future plants to ensure that the realized level of safety is commensurate with the safety analyses used in regulatory decisions.

  6. Neutron transport analysis for nuclear reactor design

    DOE Patents [OSTI]

    Vujic, Jasmina L. (Lisle, IL)

    1993-01-01T23:59:59.000Z

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values.

  7. Neutron transport analysis for nuclear reactor design

    DOE Patents [OSTI]

    Vujic, J.L.

    1993-11-30T23:59:59.000Z

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values. 28 figures.

  8. Design and control of tubular autothermal reactor: An evolutionary approach

    SciTech Connect (OSTI)

    Chylla, R.W. Jr.; Adomaitis, R.A.; Cinar, A.

    1986-01-01T23:59:59.000Z

    A systematic procedure for the design of a reactor system and controller for an autothermal process is proposed. Use of Structural Dominance Analysis eliminates the need for many detailed simulation runs to determine best reactor and controller configuration. Multistage wall-cooled reactors with cocurrent coolant are found to be a superior design. Reactor control is achieved using cold feed bypass gas to regulate in inlet temperature to each reactor bed. A controller is designed using the Internal Model Control structure. Performance and robustness are investigated using a first-order diagonal filter. 15 refs.

  9. Fusion reactor blanket-main design aspects

    SciTech Connect (OSTI)

    Strebkov, Yu.; Sidorov, A.; Danilov, I. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation)

    1994-12-31T23:59:59.000Z

    The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm{sup 2} and 7 MWa/m{sup 2} accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket.

  10. Advanced High Temperature Reactor Neutronic Core Design

    SciTech Connect (OSTI)

    Ilas, Dan [ORNL] [ORNL; Holcomb, David Eugene [ORNL] [ORNL; Varma, Venugopal Koikal [ORNL] [ORNL

    2012-01-01T23:59:59.000Z

    The AHTR is a 3400 MW(t) FHR class reactor design concept intended to serve as a central generating station type power plant. While significant technology development and demonstration remains, the basic design concept appears sound and tolerant of much of the remaining performance uncertainty. No fundamental impediments have been identified that would prevent widespread deployment of the concept. This paper focuses on the preliminary neutronic design studies performed at ORNL during the fiscal year 2011. After a brief presentation of the AHTR design concept, the paper summarizes several neutronic studies performed at ORNL during 2011. An optimization study for the AHTR core is first presented. The temperature and void coefficients of reactivity are then analyzed for a few configurations of interest. A discussion of the limiting factors due to the fast neutron fluence follows. The neutronic studies conclude with a discussion of the control and shutdown options. The studies presented confirm that sound neutronic alternatives exist for the design of the AHTR to maintain full passive safety features and reasonable operation conditions.

  11. TOKAMAK REACTOR DESIGNS AS A FUNCTION OF ASPECT RATIO

    E-Print Network [OSTI]

    California at Los Angeles, University of

    GA­A23168 TOKAMAK REACTOR DESIGNS AS A FUNCTION OF ASPECT RATIO by C.P.C. WONG and R.D. STAMBAUGH or reflect those of the United States Government or any agency thereof. #12;GA­A23168 TOKAMAK REACTOR DESIGNS JULY 1999 #12;C.P.C. WONG AND R.D. STAMBAUGH TOKAMAK REACTOR DESIGNS AS A FUNCTION OF ASPECT RATIO

  12. Advanced Modularity Design for The MIT Pebble Bed Reactor

    E-Print Network [OSTI]

    Advanced Modularity Design for The MIT Pebble Bed Reactor Andrew C. Kadak Department of Nuclear Reactor Technology Institute of Nuclear and New Energy Technology Friendship Hotel, Haidian District Beijing, China September 22-24, 2004 Abstract The future of all reactors will depend on whether they can

  13. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    OF LARGE FAST REACTORS Calculation examples A typicalMonte Carlo Reactor Physics Burnup Calculation Code. Tech.reactor core design from experience and coarse calculations

  14. Core design studies for advanced burner test reactor.

    SciTech Connect (OSTI)

    Yang, W. S.; Kim, T. K.; Hill, R. N.; Nuclear Engineering Division

    2008-01-01T23:59:59.000Z

    The U.S. government announced in February 2006 the Global Nuclear Energy Partnership (GNEP) to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. The advanced burner reactor (ABR) based on a fast spectrum is one of the three major technologies to be demonstrated in GNEP. In FY06, a pre-conceptual design study was performed to develop an advanced burner test reactor (ABTR) that supports development of a prototype full-scale ABR, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR were (1) to demonstrate reactor-based transmutation of transuranics (TRU) as part of an advanced fuel cycle, (2) to qualify the TRU-containing fuels and advanced structural materials needed for a full-scale ABR, (3) to support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. Based on these objectives, core design and fuel cycle studies were performed to develop ABTR core designs, which can accommodate the expected changes of the TRU feed and the conversion ratio. Various option and trade-off studies were performed to determine the appropriate power level and conversion ratio. Both ternary metal alloy (U-TRU-10Zr) and mixed oxide (UO{sub 2}-TRUO{sub 2}) fuel forms have been considered with TRU feeds from weapons-grade plutonium (WG-Pu) and TRU recovered from light water reactor spent fuel (LWR-SF). Reactor performances were evaluated in detail including equilibrium cycle core parameters, mass flow, power distribution, kinetic parameters, reactivity feedback coefficient, reactivity control requirements and shutdown margins, and spent fuel characteristics. Trade-off studies on power level suggested that about 250 MWt is a reasonable compromise to allow a low project cost, at the same time providing a reasonable prototypic irradiation environment for demonstrating TRU-based fuels. Preliminary design studies showed that it is feasible to design the ABTR to accommodate a wide range of conversion ratio (CR) by employing different assembly designs. The TRU enrichments required for various conversion ratios and the irradiation database suggested a phased approach with initial startup using conventional enrichment plutonium-based fuel and gradual transitioning to full core loading of transmutation fuel after its qualification phase (resulting in {approx}0.6 CR). The low CR transmutation fuel tests can be accommodated in the designated test assemblies, and if fully developed, core conversion to low CR fuel can be envisioned. Reference ABTR core designs with a rated power of 250 MWt were developed for ternary metal alloy and mixed oxide fuels based on WG-Pu feed. The reference core contains 54 driver, 6 test fuel, and 3 test material assemblies. For the startup core designs, the calculated TRU conversion ratio is 0.65 for the metal fuel core and 0.64 for the oxide fuel core. Both the metal and oxide cores show good performances. The metal fuel core requires an average TRU enrichment of 18.8% and yields a reactivity swing of 1.2 %{Delta}k over the 4-month cycle. The core average flux level is {approx}2.4 x 10{sup 15} n/cm{sup 2}s, and test assembly flux level is {approx}2.8 x 10{sup 15} n/cm{sup 2}s. Compared to the metal fuel core, the lower density oxide fuel core requires an average TRU enrichment of 21.8%, which results in a 780 kg TRU loading (as compared to 732 kg for metal) despite a {approx}9% smaller heavy metal inventory. The lower heavy metal inventory increases the burnup reactivity swing by {approx}10% and reduces the flux levels by {approx}8%. Alternative designs were also studied for a LWR-SF TRU feed and a low conversion ratio, including the recycle of the ABTR spent fuel TRU. The lower fissile contents of the LWR-SF TRU relative to the WG-Pu TRU significantly increase the required TRU enrichment of the startup cores to maintain the same cycle length. The even lower fissile fraction of the ABTR spent fuel TRU furt

  15. Novel Reactor Design and Metrology Study for Tungsten ALD process

    E-Print Network [OSTI]

    Rubloff, Gary W.

    species Viscous flow condition Short gas residence time Fast gas switching Reactant + carrier gas Multiple Operation Modes Exposure Purge Small reactor volume Throttle Valve 5 torr 10-5 Torr carrier gas 5 torr 10Novel Reactor Design and Metrology Study for Tungsten ALD process Laurent Henn-Lecordier, Wei Lei

  16. advanced reactor design: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    20 21 22 23 24 25 Next Page Last Page Topic Index 1 Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors University of California...

  17. advanced reactor designs: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    20 21 22 23 24 25 Next Page Last Page Topic Index 1 Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors University of California...

  18. Innovative fuel designs for high power density pressurized water reactor

    E-Print Network [OSTI]

    Feng, Dandong, Ph. D. Massachusetts Institute of Technology

    2006-01-01T23:59:59.000Z

    One of the ways to lower the cost of nuclear energy is to increase the power density of the reactor core. Features of fuel design that enhance the potential for high power density are derived based on characteristics of ...

  19. Design of a nuclear reactor system for lunar base applications

    E-Print Network [OSTI]

    Griffith, Richard Odell

    1986-01-01T23:59:59.000Z

    DESIGN OF A NUCLEAR REACTOR SYSTEM FOR LUNAR BASE APPLICATIONS A Thesis by RICMARD ODELL GRIFFITH Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE... August 1986 Major Subject: Nuclear Engineer ing DESIGN OF A NUCLEAR REACTOR SYSTEM FOR LUNAR BASE APPLICATIONS A Thesis by RICHARD ODELL GRIFFITH Appr oved as to style and content by: Carl A. Endman (Chain of Committee) Cer aid A. Schlapper...

  20. An Identity and Certificate Manager

    E-Print Network [OSTI]

    Wu, Brian C

    2007-01-01T23:59:59.000Z

    We have designed and implemented a software library, called Identity and Certificate Manager (ICM), for managing, using, and exchanging application-level usernames, users' digital certificates, and cryptographic ...

  1. DESIGN OF A TOKAMAK FUSION REACTOR FIRST WALL ARMOR AGAINST NEUTRAL BEAM IMPINGEMENT

    E-Print Network [OSTI]

    Myers, Richard Allen

    2011-01-01T23:59:59.000Z

    et. a1. , "A Conceptual Tokamak Reactor Design, umtAK-II,"et. al.. "A Non-Circular Tokamak 'Power Reactor Design,"Contt'ol in Near Term Tokamak Reactors," Proceedings of the

  2. Conceptual Design of a Large, Passive Pressure-Tube Light Water Reactor

    E-Print Network [OSTI]

    Hejzlar, P.

    A design for a large, passive, light water reactor has been developed. The proposed concept is a pressure tube reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated ...

  3. Design study of the deep-sea reactor X

    SciTech Connect (OSTI)

    Iida, Hiromasa (Japan Atomic Energy Research Inst., Ibaraki (Japan)); Ishizaka, Yuichi (Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)); Kim, Y.C.; Yamaguchi, Chouichi (Japan Research Inst., Ltd., Tokyo (Japan))

    1994-07-01T23:59:59.000Z

    The deep-sea reactor X (DRX) is a small nuclear plant designed to provide undersea power sources. It has the full advantages of nuclear reactors and can provide large power capacity and does not require oxygen for power production. An application conceivable in the near future is that for a submersible. The Japan Atomic Energy Research Institute is conducting a design study of a 150-kW(electric) DRX plant for a deep-sea research vessel. It has a so-called integrated pressurized water reactor,'' having a steam generator inside the reactor vessel. A pressure shell includes a turbine and a generator as well as a reactor vessel, resulting in a very compact electricity producing plant. It should be easy to operate and have high passive safety characteristics; namely, a short startup time, good reactor response to power demand changes, and passive core flooding and decay heat removal in case of an accident. Transient analyses including those for load follow-up, reactor startup, and accidents have been conducted. The results show that the DRX has excellent inherent characteristics satisfying those requirements.

  4. A 5 MW TRIGA reactor design for radioisotope production

    SciTech Connect (OSTI)

    Veca, Anthony R.; Whittemore, William L. [General Atomics, San Diego, CA (United States)

    1994-07-01T23:59:59.000Z

    The production and preparation of commercial-scale quantities of radioisotopes has become an important activity as their medical and industrial applications continue to expand. There are currently various large multipurpose research reactors capable of producing ample quantities of radioisotopes. These facilities, however, have many competing demands placed upon them by a wide variety of researchers and scientific programs which severely limit their radioisotope production capability. A demonstrated need has developed for a simpler reactor facility dedicated to the production of radioisotopes on a commercial basis. This smaller, dedicated reactor could provide continuous fission and activation product radioisotopes to meet commercial requirements for the foreseeable future. The design of a 5 MW TRIGA reactor facility, upgradeable to 10 MW, dedicated to the production of industrial and medical radioisotopes is discussed. A TRIGA reactor designed specifically for this purpose with its demonstrated long core life and simplicity of operation would translate into increased radioisotope production. As an example, a single TRIGA could supply the entire US needs for Mo-99. The facility is based on the experience gained by General Atomics in the design, installation, and construction of over 60 other TRIGAs over the past 35 years. The unique uranium-zirconium hydride fuel makes TRIGA reactors inexpensive to build and operate, reliable in their simplicity, highly flexible due to unique passive safety, and environmentally friendly because of minimal power requirements and long-lived fuel. (author)

  5. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    4 Reactivity feedback of large fast reactors 4.1temperature . . . . . . . . . . . . . . . . . . Fast reactorfission gas plenum212 Conventional fast reactor core design

  6. Westinghouse Small Modular Reactor nuclear steam supply system design

    SciTech Connect (OSTI)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01T23:59:59.000Z

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)

  7. Basis for NGNP Reactor Design Down-Selection

    SciTech Connect (OSTI)

    L.E. Demick

    2010-08-01T23:59:59.000Z

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  8. Basis for NGNP Reactor Design Down-Selection

    SciTech Connect (OSTI)

    L.E. Demick

    2011-11-01T23:59:59.000Z

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  9. Preliminary design studies on the Broad Application Test Reactor

    SciTech Connect (OSTI)

    Terry, W.J. [ed.; Terry, W.K.; Ryskamp, J.M.; Jahshan, S.N.; Fletcher, C.D.; Moore, R.L.; Leyse, C.F.; Ottewitte, E.H.; Motloch, C.G.; Lacy, J.M.

    1992-08-01T23:59:59.000Z

    This report describes progress made at the Idaho National Engineering Laboratory during the first three quarters of Fiscal Year (FY) 1992 on the Laboratory-Directed Research and Development (LDRD) project to perform preliminary design studies on the Broad Application Test Reactor (BATR). This work builds on the FY-92 BATR studies, which identified anticipated mission and safety requirements for BATR and assessed a variety of reactor concepts for their potential capability to meet those requirements. The main accomplishment of the FY-92 BATR program is the development of baseline reactor configurations for the two conventional conceptual test reactors recommended in the FY-91 report. Much of the present report consists of descriptions and neutronics and thermohydraulics analyses of these baseline configurations. In addition, we considered reactor safety issues, compared the consequences of steam explosions for alternative conventional fuel types, explored a Molten Chloride Fast Reactor concept as an alternate BATR design, and examined strategies for the reduction of operating costs. Work planned for the last quarter of FY-92 is discussed, and recommendations for future work are also presented.

  10. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    fission gas plenum212 Conventional fast reactor core designGUPTA. “A Compact Gas-Cooled Fast Reactor with an Ultra-Longbreed and burn gas-cooled fast reactor”. Ph.D. Thesis. MIT,

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

    SciTech Connect (OSTI)

    S. Blaine Grover

    2005-10-01T23:59:59.000Z

    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.

  12. B.F.A. in Studio Art Graduation Certification Fabric Design Area of Emphasis Student Name _________________________ SS #______-_____-_______ rev. 04/12

    E-Print Network [OSTI]

    Arnold, Jonathan

    you can graduate. ARST 4710 BFA Project in Fabric Design __3__ _____ __________________ ExhibitionB.F.A. in Studio Art Graduation Certification Fabric Design Area of Emphasis Student Name (542-1522) for a graduation certification appointment at least 2 terms before graduation

  13. Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency

    SciTech Connect (OSTI)

    R. Wigeland; K. Hamman

    2009-09-01T23:59:59.000Z

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving greater thermal efficiency, since it causes the fuel pins in the center of the subassembly to operate at higher temperatures than those near the hexcan walls, and it is the temperature limit(s) for those fuel pins that limits the average coolant outlet temperature. Fuel subassembly design changes are being investigated using computational fluid dynamics (CFD) to quantify the effect that the design changes have on reducing the intra-subassembly coolant flow and temperature distribution. Simulations have been performed for a 19-pin test subassembly geometry using typical fuel pin diameters and wire wrap spacers. The results have shown that it may be possible to increase the average coolant outlet temperature by 20 C or more without changing the peak temperatures within the subassembly. These design changes should also be effective for reactor designs using subassemblies with larger numbers of fuel pins. R. Wigeland, Idaho National Laboratory, P.O. Box 1625, Mail Stop 3860, Idaho Falls, ID, U.S.A., 83415-3860 email – roald.wigeland@inl.gov fax (U.S.) – 208-526-2930

  14. Reactor core design and modeling of the MIT research reactor for conversion to LEU

    SciTech Connect (OSTI)

    Newton, Thomas H. Jr. [Nuclear Reactor Laboratory, Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Olson, Arne P.; Stillman, John A. [RERTR Program, Argonne National Laboratory, Argonne, IL 60439 (United States)

    2008-07-15T23:59:59.000Z

    Feasibility design studies for conversion of the MIT Research Reactor (MITR) to LEU are described. Because the reactor fuel has a rhombic cross section, a special input processor was created in order to model the reactor in great detail with the REBUS-PC diffusion theory code, in 3D (triangular-z) geometry. Comparisons are made of fuel assembly power distributions and control blade worth vs. axial position, between REBUS-PC results and Monte Carlo predictions from the MCNP code. Results for the original HEU core at zero burnup are also compared with measurement. These two analysis methods showed remarkable agreement. Ongoing fuel cycle studies are summarized. A status report will be given as to results thus far that affect key design decisions. Future work plans and schedules to achieve completion of the conversion are presented. (author)

  15. Core design and reactor physics of a breed and burn gas-cooled fast reactor

    E-Print Network [OSTI]

    Yarsky, Peter

    2005-01-01T23:59:59.000Z

    In order to fulfill the goals set forth by the Generation IV International Forum, the current NERI funded research has focused on the design of a Gas-cooled Fast Reactor (GFR) operating in a Breed and Burnm (B&B) fuel cycle ...

  16. NGNP Nuclear-Industrial Facility and Design Certification Boundaries White Paper

    SciTech Connect (OSTI)

    Thomas E. Hicks

    2011-07-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) Project was initiated at Idaho National Laboratory by the U.S. Department of Energy pursuant to the 2005 Energy Policy Act and based on research and development activities supported by the Generation IV Nuclear Energy Systems Initiative. The principal objective of the NGNP Project is to support commercialization of the high temperature gas-cooled reactor (HTGR) technology. The HTGR is helium cooled and graphite moderated and can operate at reactor outlet temperatures much higher than those of conventional light water reactor (LWR) technologies. Accordingly, it can be applied in many industrial applications as a substitute for burning fossil fuels, such as natural gas, in addition to producing electricity, which is the principal application of current LWRs. These varied industrial applications may involve a standard HTGR modular design using different Energy Conversion Systems. Additionally, some of these process heat applications will require process heat delivery systems to lie partially outside the HTGR operator’s facility.

  17. ChBE 4300 Kinetics and Reactor Design (required course) Credit: 3-0-3

    E-Print Network [OSTI]

    Sherrill, David

    , and (ii) reactor design for the homogeneous reaction systems. The design principles for ideal homogeneousChBE 4300 Kinetics and Reactor Design (required course) Credit: 3-0-3 Prerequisite in terms of reaction mechanisms, kinetics, and reactor design. Both homogeneous and heterogeneous reactions

  18. The occupational endorsement certification in welding and materials technology is designed for beginning students who want to learn to

    E-Print Network [OSTI]

    Ickert-Bond, Steffi

    WELDING & MATERIALS TECHNOLOGY The occupational endorsement certification in welding and materials technology is designed for beginning students who want to learn to cut and weld, as well as students seeking more advanced courses who would like to take their welding skills further. Our program emphasizes

  19. Design of passive decay heat removal system for the lead cooled flexible conversion ratio fast reactor

    E-Print Network [OSTI]

    Whitman, Joshua (Joshua J.)

    2007-01-01T23:59:59.000Z

    The lead-cooled flexible conversion ratio fast reactor shows many benefits over other fast-reactor designs; however, the higher power rating and denser primary coolant present difficulties for the design of a passive decay ...

  20. Thermal hydraulic design of a salt-cooled highly efficient environmentally friendly reactor

    E-Print Network [OSTI]

    Whitman, Joshua (Joshua J.)

    2009-01-01T23:59:59.000Z

    A 1 OOOMWth liquid-salt cooled thermal spectrum reactor was designed with a long fuel cycle, and high core exit temperature. These features are desirable in a reactor designed to provide process heat applications such as ...

  1. System 80+{trademark} standard design: CESSAR design certification. Volume 15: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{trademark} Standard Design.

  2. System 80+{trademark} standard design: CESSAR design certification. Volume 14: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{trademark} Standard Design.

  3. Influence of modeling and simulation on the maturation of plasma technology: Feature evolution and reactor design

    E-Print Network [OSTI]

    Kushner, Mark

    requires fluxes from reactor scale phenom- ena. To achieve the goal of using MS for first principles design and reactor design David B. Gravesa) Department of Chemical Engineering, University of California, Berkeley and future potential of MS for feature evolution and plasma reactor design. © 2003 American Vacuum Society

  4. Advanced Neutron Source reactor control and plant protection systems design

    SciTech Connect (OSTI)

    Anderson, J.L.; Battle, R.E.; March-Leuba, J. (Oak Ridge National Lab., TN (United States)); Khayat, M.I. (Tennessee Univ., Knoxville, TN (United States))

    1992-01-01T23:59:59.000Z

    This paper describes the reactor control and plant protection systems' conceptual design of the Advanced Neutron Source (ANS). The Plant Instrumentation, Control, and Data Systems and the Reactor Instrumentation and Control System of the ANS are planned as an integrated digital system with a hierarchical, distributed control structure of qualified redundant subsystems and a hybrid digital/analog protection system to achieve the necessary fast response for critical parameters. Data networks transfer information between systems for control, display, and recording. Protection is accomplished by the rapid insertion of negative reactivity with control rods or other reactivity mechanisms to shut down the fission process and reduce heat generation in the fuel. The shutdown system is designed for high functional reliability by use of conservative design features and a high degree of redundance and independence to guard against single failures. Two independent reactivity control systems of different design principles are provided, and each system has multiple independent rods or subsystems to provide appropriate margin for malfunctions such as stuck rods or other single failures. Each system is capable of maintaining the reactor in a cold shutdown condition independently of the functioning of the other system. A highly reliable, redundant channel control system is used not only to achieve high availability of the reactor, but also to reduce challenges to the protection system by maintaining important plant parameters within appropriate limits. The control system has a number of contingency features to maintain acceptable, off-normal conditions in spite of limited control or plant component failures thereby further reducing protection system challenges.

  5. System 80+{trademark} standard design: CESSAR design certification. Volume 5: Amendment I

    SciTech Connect (OSTI)

    Not Available

    1990-12-21T23:59:59.000Z

    This report has been prepared in support of the industry effort to standardize nuclear plant designs. The documents in this series describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design.

  6. Energy Department Issues Green Building Certification System...

    Office of Environmental Management (EM)

    Issues Green Building Certification System Final Rule to Support Increased Energy Measurement and Efficient Building Design Energy Department Issues Green Building Certification...

  7. Simulation-Based Design and Experimental Evaluation of a Spatially Controllable CVD Reactor

    E-Print Network [OSTI]

    Rubloff, Gary W.

    Simulation-Based Design and Experimental Evaluation of a Spatially Controllable CVD Reactor Jae CVD reactor system has been developed that can explicitly control the spatial profile of gas, opening the door to a new class of flexible and highly controllable CVD reactor designs. © 2005 American

  8. Effects of an Advanced Reactor’s Design, Use of Automation, and Mission on Human Operators

    SciTech Connect (OSTI)

    Jeffrey C. Joe; Johanna H. Oxstrand

    2014-06-01T23:59:59.000Z

    The roles, functions, and tasks of the human operator in existing light water nuclear power plants (NPPs) are based on sound nuclear and human factors engineering (HFE) principles, are well defined by the plant’s conduct of operations, and have been validated by years of operating experience. However, advanced NPPs whose engineering designs differ from existing light-water reactors (LWRs) will impose changes on the roles, functions, and tasks of the human operators. The plans to increase the use of automation, reduce staffing levels, and add to the mission of these advanced NPPs will also affect the operator’s roles, functions, and tasks. We assert that these factors, which do not appear to have received a lot of attention by the design engineers of advanced NPPs relative to the attention given to conceptual design of these reactors, can have significant risk implications for the operators and overall plant safety if not mitigated appropriately. This paper presents a high-level analysis of a specific advanced NPP and how its engineered design, its plan to use greater levels of automation, and its expanded mission have risk significant implications on operator performance and overall plant safety.

  9. Assessment of innovative fuel designs for high performance light water reactors

    E-Print Network [OSTI]

    Carpenter, David Michael

    2006-01-01T23:59:59.000Z

    To increase the power density and maximum allowable fuel burnup in light water reactors, new fuel rod designs are investigated. Such fuel is desirable for improving the economic performance light water reactors loaded with ...

  10. Reactor and shielding design implications of clustering nuclear thermal rockets

    SciTech Connect (OSTI)

    Buksa, J.J.; Houts, M.G. (Los Alamos National Laboratory, NM (United States))

    1992-07-01T23:59:59.000Z

    This paper examines design considerations in the context of engine-out accidents in clustered nuclear-thermal rocket stages, and an accident-management protocol is devised. Safety and performance issues are considered in the light of designs for the reactor and shielding elements of ROVER/NERVA-type engines. The engine-out management process involves: phase one, in which sufficient propulsive power is guaranteed for mission completion; and phase two, in which engine failure is isolated and not allowed to propagate to other engines or to the spacecraft. Phase-one designs can employ spare engines, throttled engines, and/or long-burning engines. Phase-two safety concepts can include techniques for cooling or jettisoning the failed engines. Engine-out management philosophies are shown to be shaped by a combination of safety and mission-trajectory requirements. 6 refs.

  11. Fast reactor power plant design having heat pipe heat exchanger

    DOE Patents [OSTI]

    Huebotter, P.R.; McLennan, G.A.

    1984-08-30T23:59:59.000Z

    The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

  12. Fast reactor power plant design having heat pipe heat exchanger

    DOE Patents [OSTI]

    Huebotter, Paul R. (Western Springs, IL); McLennan, George A. (Downers Grove, IL)

    1985-01-01T23:59:59.000Z

    The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

  13. The design of a reduced diameter Pebble Bed Modular Reactor for reactor pressure vessel transport by railcar

    E-Print Network [OSTI]

    Everson, Matthew S

    2009-01-01T23:59:59.000Z

    Many desirable locations for Pebble Bed Modular Reactors are currently out of consideration as construction sites for current designs due to limitations on the mode of transportation for large RPVs. In particular, the ...

  14. Building Operator Certification: Improving Commercial Building Energy Efficiency Through Operator Training and Certification

    E-Print Network [OSTI]

    Putnam, C.; Mulak, A.

    2001-01-01T23:59:59.000Z

    Building Operator Certification (BOC) is a competency-based certification for building operators designed to improve the energy efficiency of commercial buildings. Operators earn certification by attending training sessions and completing project...

  15. 22.39 Integration of Reactor Design, Operations, and Safety, Fall 2005

    E-Print Network [OSTI]

    Todreas, Neil E.

    This course integrates studies of reactor physics and engineering sciences into nuclear power plant design. Topics include materials issues in plant design and operations, aspects of thermal design, fuel depletion and ...

  16. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    SciTech Connect (OSTI)

    Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

    1980-09-01T23:59:59.000Z

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb/sub 3/Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered.

  17. Controller design issues in the feedback control of radio frequency plasma processing reactors

    E-Print Network [OSTI]

    Kushner, Mark

    Controller design issues in the feedback control of radio frequency plasma processing reactors feedback control of inductively coupled plasma processing reactors for polysilicon etching and be successfully used for feedback control of plasma processing reactors.4 There are many control strate- gies

  18. Robust controller design for temperature tracking problems in jacketed batch reactors

    E-Print Network [OSTI]

    Palanki, Srinivas

    Robust controller design for temperature tracking problems in jacketed batch reactors Vishak for temperature tracking problems in batch reactors in the presence of parametric uncertainty. The controller has]. Control is achieved by manipulating the heat content from the jacket to the reactor. In the past

  19. OPTIMAL DESIGN OF A HIGH PRESSURE ORGANOMETALLIC CHEMICAL VAPOR DEPOSITION REACTOR

    E-Print Network [OSTI]

    OPTIMAL DESIGN OF A HIGH PRESSURE ORGANOMETALLIC CHEMICAL VAPOR DEPOSITION REACTOR K.J. BACHMANN vapor deposition (HPOMCVD) reactor for use in thin film crystal growth. The advantages of such a reactor decomposition pressures and increased control over local stoichiometry and defect formation. While we focus here

  20. Improved Design of Nuclear Reactor Control System | U.S. DOE...

    Office of Science (SC) Website

    Improved Design of Nuclear Reactor Control System Nuclear Physics (NP) NP Home About Research Facilities Science Highlights Benefits of NP Applications of Nuclear Science...

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

    SciTech Connect (OSTI)

    Pan, Paul Y [Los Alamos National Laboratory

    2010-12-10T23:59:59.000Z

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

  2. One pass core design of a super fast reactor

    SciTech Connect (OSTI)

    Liu, Qingjie; Oka, Yoshiaki [Cooperative Major in Nuclear Energy, Waseda University, Tokyo 169-8555 (Japan)

    2013-07-01T23:59:59.000Z

    One pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern can significantly simplify the core concept. Upper core structure, coolant flow scheme as well as refueling procedure are as simple as in PWR. In one pass core design, supercritical-pressure water is at approximately 25.0 MPa and enters the core at 280 C. degrees and is heated up in one through flow pattern upwardly to the average outlet temperature of 500 C. degrees. Great density change in vertical direction can cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO{sub 2} fuel blanket assemblies also introduces great power increase during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at the beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of the blanket assembly. This can help to mitigate the power change in blanket due to Pu accumulation and to increase the outlet temperature of the blanket during cycle. Neutron transport and thermohydraulics coupled calculation shows that this design can satisfy the requirement in the Super FR principle for both 500 C. degrees outlet temperature and negative coolant void reactivity. (authors)

  3. High Flux Isotope Reactor cold neutron source reference design concept

    SciTech Connect (OSTI)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01T23:59:59.000Z

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  4. Conceptual design of an annular-fueled superheat boiling water reactor

    E-Print Network [OSTI]

    Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

    2011-01-01T23:59:59.000Z

    The conceptual design of an annular-fueled superheat boiling water reactor (ASBWR) is outlined. The proposed design, ASBWR, combines the boiler and superheater regions into one fuel assembly. This ensures good neutron ...

  5. High flux isotope reactor cold source preconceptual design study report

    SciTech Connect (OSTI)

    Selby, D.L.; Bucholz, J.A.; Burnette, S.E. [and others

    1995-12-01T23:59:59.000Z

    In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH{sub 2} moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project.

  6. Boiling water neutronic reactor incorporating a process inherent safety design

    DOE Patents [OSTI]

    Forsberg, Charles W. (Kingston, TN)

    1987-01-01T23:59:59.000Z

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

  7. Boiling water neutronic reactor incorporating a process inherent safety design

    DOE Patents [OSTI]

    Forsberg, C.W.

    1985-02-19T23:59:59.000Z

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

  8. A brief history of design studies on innovative nuclear reactors

    SciTech Connect (OSTI)

    Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com [Emeritus Professor, Tokyo Institute of Technology (Japan)

    2014-09-30T23:59:59.000Z

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  9. Nuclear facilities: criteria for the design and operation of ventilation systems for nuclear installations other than nuclear reactors

    E-Print Network [OSTI]

    International Organization for Standardization. Geneva

    2004-01-01T23:59:59.000Z

    Nuclear facilities: criteria for the design and operation of ventilation systems for nuclear installations other than nuclear reactors

  10. A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

    SciTech Connect (OSTI)

    John Darrell Bess

    2008-06-01T23:59:59.000Z

    A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.

  11. Computational Science and Engineering Certification for Atmospheric Sciences Undergraduate Students

    E-Print Network [OSTI]

    Gilbert, Matthew

    Computational Science and Engineering Certification for Atmospheric Sciences Undergraduate Students The Computational Science and Engineering certificate program is designed to provide ATMS under- graduate students a certificate in "Computational Science and Engineering," students must complete the required courses listed

  12. Safety aspects of the US advanced LMR (liquid metal reactor) design

    SciTech Connect (OSTI)

    Pedersen, D.R.; Gyorey, G.L.; Marchaterre, J.F.; Rosen, S. (Argonne National Lab., IL (USA); General Electric Co., San Jose, CA (USA); Argonne National Lab., IL (USA); USDOE Assistant Secretary for Nuclear Energy, Washington, DC (USA))

    1989-01-01T23:59:59.000Z

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. This paper discusses the US regulatory framework for design of an ALMR, safety aspects of the IFR program at ANL, the IFR fuel cycle and actinide recycle, and the ALMR plant design program at GE. 6 refs., 5 figs.

  13. Conceptual design of a new homogeneous reactor for medical radioisotope Mo-99/Tc-99m production

    SciTech Connect (OSTI)

    Liem, Peng Hong [Nippon Advanced Information Service (NAIS Co., Inc.) Scientific Computational Division, 416 Muramatsu, Tokaimura, Ibaraki (Japan); Tran, Hoai Nam [Chalmers University of Technology, Dept. of Applied Physics, Div. of Nuclear Engineering, SE-412 96 Gothenburg (Sweden); Sembiring, Tagor Malem [National Nuclear Energy Agency (BATAN), Center for Reactor Technology and Nuclear Safety, Kawasan Puspiptek, Serpong, Tangerang Selatan, Banten (Indonesia); Arbie, Bakri [PT MOTAB Technology, Kedoya Elok Plaza Blok DA 12, Jl. Panjang, Kebun Jeruk, Jakarta Barat (Indonesia)

    2014-09-30T23:59:59.000Z

    To partly solve the global and regional shortages of Mo-99 supply, a conceptual design of a nitrate-fuel-solution based homogeneous reactor dedicated for Mo-99/Tc-99m medical radioisotope production is proposed. The modified LEU Cintichem process for Mo-99 extraction which has been licensed and demonstrated commercially for decades by BATAN is taken into account as a key design consideration. The design characteristics and main parameters are identified and the advantageous aspects are shown by comparing with the BATAN's existing Mo-99 supply chain which uses a heterogeneous reactor (RSG GAS multipurpose reactor)

  14. Linear Parameter-Varying versus Linear Time-Invariant Control Design for a Pressurized Water Reactor

    E-Print Network [OSTI]

    Bodenheimer, Bobby

    -dependent control to a nuclear pressurized water reactor is investigated and is compared to that of using an H1Linear Parameter-Varying versus Linear Time-Invariant Control Design for a Pressurized Water Reactor Pascale Bendotti y Electricit e de France Direction des Etudes et Recherches 6 Quai Watier, 78401

  15. Reactor Design and Decommissioning - An Overview of International Activities in Post Fukushima Era1 - 12396

    SciTech Connect (OSTI)

    Devgun, Jas S. [Nuclear Power Technologies, Sargent and Lundy LLC, Chicago, IL (United States); Laraia, Michele [private consultant, formerly from IAEA, Kolonitzgasse 10/2, 1030, Vienna (Austria); Pescatore, Claudio [OECD, Nuclear Energy Agency, Issy-les-Moulineaux, Paris (France); Dinner, Paul [International Atomic Energy Agency, Wagramerstrasse 5, A-1400 Vienna (Austria)

    2012-07-01T23:59:59.000Z

    Accidents at the Fukushima Dai-ichi reactors as a result of the devastating earthquake and tsunami of March 11, 2011 have not only dampened the nuclear renaissance but have also initiated a re-examination of the design and safety features for the existing and planned nuclear reactors. Even though failures of some of the key site features at Fukushima can be attributed to events that in the past would have been considered as beyond the design basis, the industry as well as the regulatory authorities are analyzing what features, especially passive features, should be designed into the new reactor designs to minimize the potential for catastrophic failures. It is also recognized that since the design of the Fukushima BWR reactors which were commissioned in 1971, many advanced safety features are now a part of the newer reactor designs. As the recovery efforts at the Fukushima site are still underway, decisions with respect to the dismantlement and decommissioning of the damaged reactors and structures have not yet been finalized. As it was with Three Mile Island, it could take several decades for dismantlement, decommissioning and clean up, and the project poses especially tough challenges. Near-term assessments have been issued by several organizations, including the IAEA, the USNRC and others. Results of such investigations will lead to additional improvements in system and site design measures including strengthening of the anti-tsunami defenses, more defense-in-depth features in reactor design, and better response planning and preparation involving reactor sites. The question also arises what would the effect be on the decommissioning scene worldwide, and what would the effect be on the new reactors when they are eventually retired and dismantled. This paper provides an overview of the US and international activities related to recovery and decommissioning including the decommissioning features in the reactor design process and examines these from a new perspective in the post Fukushima -accident era. Accidents at the Fukushima Daiichi reactors in the aftermath of the devastating earthquake and tsunami of March 11, 2011 have slowed down the nuclear renaissance world-wide and may have accelerated decommissioning either because some countries have decided to halt or reduce nuclear, or because the new safety requirements may reduce life-time extensions. Even in countries such as the UK and France that favor nuclear energy production existing nuclear sites are more likely to be chosen as sites for future NPPs. Even as the site recovery efforts continue at Fukushima and any decommissioning decisions are farther into the future, the accidents have focused attention on the reactor designs in general and specifically on the Fukushima type BWRs. The regulatory authorities in many countries have initiated a re-examination of the design of the systems, structures and components and considerations of the capability of the station to cope with beyond-design basis events. Enhancements to SSCs and site features for the existing reactors and the reactors that will be built will also impact the decommissioning phase activities. The newer reactor designs of today not only have enhanced safety features but also take into consideration the features that will facilitate future decommissioning. Lessons learned from past management and operation of reactors as well as the lessons from decommissioning are incorporated into the new designs. However, in the post-Fukushima era, the emphasis on beyond-design-basis capability may lead to significant changes in SSCs, which eventually will also have impact on the decommissioning phase. Additionally, where some countries decide to phase out the nuclear power, many reactors may enter the decommissioning phase in the coming decade. While the formal updating and expanding of existing guidance documents for accident cleanup and decommissioning would benefit by waiting until the Fukushima project has progressed sufficiently for that experience to be reliably interpreted, the development of structured on-li

  16. Power conversion system design for supercritical carbon dioxide cooled indirect cycle nuclear reactors

    E-Print Network [OSTI]

    Gibbs, Jonathan Paul

    2008-01-01T23:59:59.000Z

    The supercritical carbon dioxide (S-CO?) cycle is a promising advanced power conversion cycle which couples nicely to many Generation IV nuclear reactors. This work investigates the power conversion system design and ...

  17. Conversion of methanol to light olefins on SAPO-34: kinetic modeling and reactor design

    E-Print Network [OSTI]

    Al Wahabi, Saeed M. H.

    2005-02-17T23:59:59.000Z

    design of an MTO reactor, accounting for the strong exothermicity of the process. Multi-bed adiabatic and fluidized bed technologies show good potential for the industrial process for the conversion of methanol into olefins....

  18. Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor

    E-Print Network [OSTI]

    Bean, Malcolm K.

    2011-08-01T23:59:59.000Z

    Molten salt is a promising coolant candidate for Advanced High Temperature Reactor (AHTR) Gen-IV designs. The low neutron absorption, high thermal capacity, chemical inertness, and high boiling point at low pressure of ...

  19. Design, construction and evaluation of a facility for the simulation of fast reactor blankets

    E-Print Network [OSTI]

    Forbes, Ian Alexander

    1970-01-01T23:59:59.000Z

    A facility has been designed and constructed at the MIT Reactor for the experimental investigation of typical LMFBR breeding blankets. A large converter assembly, consisting of a 20-cm-thick layer of graphite followed by ...

  20. Aerosol engineering: design and stability of aerosol reactors

    SciTech Connect (OSTI)

    Pratsinis, S.E.

    1985-01-01T23:59:59.000Z

    A theoretical study of the performance of aerosol reactors is presented. The goals of this study are (1) to identify the appropriate reactor types (batch, CSTR, and tubular) for production of aerosol with specific properties (for example, uniform size particles, high aerosol surface area, etc.) and (2) to investigate the effect of various process parameters on product aerosol characteristics and on the stability of operation of aerosol reactors. In all the reactors considered, the aerosol dynamics were detemined by chemical reaction, nucleation, and aerosol growth in the free molecule regime in the absence of coagulation at isothermal conditions. Formulation of the aerosol dynamics in terms of moments of the aerosol size distribution facilitated the numerical solution of the resulting systems of ordinary or partial differential equations. The stability characteristics of a continuous stirred tank aerosol reactor (CSTAR) were investigated since experimental data in the literature indicate that under certain conditions this reactor exhibits oscillatory behavior with respect to product aerosol concentration and size distribution.

  1. The design and construction of a 10-amplifier analog computer with provisions for nuclear reactor simulation

    E-Print Network [OSTI]

    Cox, James Robert

    1959-01-01T23:59:59.000Z

    THE DESIGN AND CONSTRUCTION OF A 10-AMPLIFIER ANALOG COMPUTER WITH PROVISIONS FOR NUCLEAR REACTOR SIMULATION A Thesis by James Robert Cox Submitted to the Graduate School of the Agricultural and Mechanical College of Texas in partial... fulfillment of the requirements for the degree of MASTER OF SCIENCE August 1959 Major Subject: Ele ctr ical Engines r ing THE DESIGN AND CONSTRUCTION OF A 10-AMPLIFIER ANALOG COMPUTER WITH PROVISIONS FOR NUCLEAR REACTOR SIMULATION A The s is by Jame...

  2. GRADUATE CERTIFICATE PROGRAM GEOGRAPHIC INFORMATION SCIENCES

    E-Print Network [OSTI]

    Doyle, Martin

    GRADUATE CERTIFICATE PROGRAM GEOGRAPHIC INFORMATION SCIENCES Department of Geography University Information Sciences (GIScience) study geospatial phenomena using an integrated set of spatial digital Certificate Program in Geographic Information Sciences offered within the Department of Geography is designed

  3. CONCEPTUAL DESIGN OF A LUNAR REGOLITH CLUSTERED-REACTOR SYSTEM

    SciTech Connect (OSTI)

    John Darrell Bess

    2009-06-01T23:59:59.000Z

    It is proposed that a fast-fission, heatpipe-cooled, lunar-surface power reactor system be divided into subcritical units that could be launched safely without the incorporation of additional spectral shift absorbers or other complex means of control. The reactor subunits are to be emplaced directly into the lunar regolith utilizing the regolith not just for shielding but as the reflector material to increase the neutron economy of the system. While a single subunit cannot achieve criticality by itself, coordinated placement of additional subunits will provide a critical reactor system for lunar surface power generation. A lunar regolith clustered-reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of a slight increase in launch mass per rated power level and an overall reduction in neutron economy when compared to a single-reactor system. Additional subunits may be launched with future missions to increase the cluster size and power according to desired lunar base power demand and lifetime. The results address the potential uncertainties associated with the lunar regolith material and emplacement of the subunit systems. Physical distance between subunits within the clustered emplacement exhibits the most significant feedback regarding changes in overall system reactivity. Narrow, deep holes will be the most effective in reducing axial neutron leakage from the core. The variation in iron concentration in the lunar regolith can directly influence the overall system reactivity although its effects are less than the more dominant factors of subunit emplacement.

  4. Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

    SciTech Connect (OSTI)

    Takeda, T.; Shimazu, Y.; Hibi, K.; Fujimura, K. [Research Inst. of Nuclear Engineering, Univ. of Fukui, 1cho-me 2gaiku 4, Kanawa-cho, Tsuruga-shi, Fukui 914-0055 (Japan)

    2012-07-01T23:59:59.000Z

    Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of this project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)

  5. Design optimization analysis of the new SPR III-M reactor

    SciTech Connect (OSTI)

    Miller, J.D.

    1993-12-31T23:59:59.000Z

    This report discusses the finite element method analysis which was used to refine the SPR III-M reactor fuel assembly mechanical design to withstand the stresses and strains of pulse-mode operation, which induces thermal shock loading in the fuel assembly components. The original reactor design was analyzed for its structural response to separate pulses at increasingly severe levels. Subsequent calculations at one consistent pulse level examined several design modifications, which will result in a significant reduction in stress in the final design.

  6. Design of a low enrichment, enhanced fast flux core for the Massachusetts Institute of Technology Research Reactor

    E-Print Network [OSTI]

    Ellis, Tyler Shawn

    2009-01-01T23:59:59.000Z

    Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating ...

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

    SciTech Connect (OSTI)

    S. Blaine Grover

    2006-10-01T23:59:59.000Z

    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.

  8. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect (OSTI)

    B. Boer; A. M. Ougouag

    2010-09-01T23:59:59.000Z

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

  9. Quality Counts Certificate

    E-Print Network [OSTI]

    Chilek, Kevin

    2007-01-31T23:59:59.000Z

    This certificate accompanies the Quality Counts program. 1 page. It comes in packages of 25 certificates....

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

    SciTech Connect (OSTI)

    Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

    2008-04-23T23:59:59.000Z

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

  11. Overall plant design specification Modular High Temperature Gas-cooled Reactor. Revision 9

    SciTech Connect (OSTI)

    NONE

    1990-05-01T23:59:59.000Z

    Revision 9 of the ``Overall Plant Design Specification Modular High Temperature Gas-Cooled Reactor,`` DOE-HTGR-86004 (OPDS) has been completed and is hereby distributed for use by the HTGR Program team members. This document, Revision 9 of the ``Overall Plant Design Specification`` (OPDS) reflects those changes in the MHTGR design requirements and configuration resulting form approved Design Change Proposals DCP BNI-003 and DCP BNI-004, involving the Nuclear Island Cooling and Spent Fuel Cooling Systems respectively.

  12. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    SciTech Connect (OSTI)

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12T23:59:59.000Z

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  13. Design of a 2400MW liquid-salt cooled flexible conversion ratio reactor

    E-Print Network [OSTI]

    Petroski, Robert C

    2008-01-01T23:59:59.000Z

    A 2400MWth liquid-salt cooled flexible conversion ratio reactor was designed, utilizing the ternary chloride salt NaCl-KCl-MgCI2 (30%-20%-50%) as coolant. The reference design uses a wire-wrapped, hex lattice core, and is ...

  14. Professional Certificate in Risk Management

    E-Print Network [OSTI]

    Carleton University

    Professional Certificate in Risk Management 2010 Program 3 Part-time Courses CRM01: Essentials of Risk Management ­ Next Session Fall 2010 CRM02: Risk Control ­ Jan. 12 - April 26, 2010 CRM03: Risk to apply for the Canadian Risk Management (CRM) designation. #12;Professional Certificate in Risk

  15. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    SciTech Connect (OSTI)

    Trond Bjornard; John Hockert

    2011-08-01T23:59:59.000Z

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC&A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC&A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC&A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR [Pty] and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC&A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR&D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present work summarized in this report, namely the development of guidance to support the consideration of MC&A in the design of both pebble-bed and prismatic-fueled HTGRs. The objective is to identify and incorporate design features into the facility design that will cost effectively aid in making MC&A more effective and efficient, with minimum impact on operations. The theft of nuclear material is addressed through both MC&A and physical protection, while the threat of sabotage is addressed principally through physical protection.

  16. Risk-informed design guidance for a Generation-IV gas-cooled fast reactor emergency core cooling system

    E-Print Network [OSTI]

    Delaney, Michael J. (Michael James), 1979-

    2004-01-01T23:59:59.000Z

    Fundamental objectives of sustainability, economics, safety and reliability, and proliferation resistance, physical protection and stakeholder relations must be considered during the design of an advanced reactor. However, ...

  17. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  18. Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

    SciTech Connect (OSTI)

    Beyer, Brian David [Los Alamos National Laboratory; Beddingfield, David H [Los Alamos National Laboratory; Durst, Philip [INL; Bean, Robert [INL

    2010-01-01T23:59:59.000Z

    The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguards criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.

  19. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    SciTech Connect (OSTI)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01T23:59:59.000Z

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

  20. Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

    SciTech Connect (OSTI)

    Koenig, D.R.; Gido, R.G.; Brandon, D.I.

    1985-01-01T23:59:59.000Z

    The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations.

  1. Thermal Response of the Hybrid Loop-Pool Design for Sodium Cooled Faster Reactors

    SciTech Connect (OSTI)

    Zhang, Hongbin; Zhao, Haihua; Davis, Cliff

    2008-09-01T23:59:59.000Z

    An innovative hybrid loop-pool design for the sodium cooled fast reactor (SFR) has been recently proposed with the primary objective of achieving cost reduction and safety enhancement. With the hybrid loop-pool design, closed primary loops are immersed in a secondary buffer tank. This design takes advantage of features from conventional both pool and loop designs to further improve economics and safety. This paper will briefly introduce the hybrid loop-pool design concept and present the calculated thermal responses for unproctected (without reactor scram) loss of forced circulation (ULOF) transients using RELAP5-3D. The analyses examine both the inherent reactivity shutdown capability and decay heat removal performance by passive safety systems.

  2. Validation of FSP Reactor Design with Sensitivity Studies of Beryllium-Reflected Critical Assemblies

    SciTech Connect (OSTI)

    John D. Bess; Margaret A. Marshall

    2013-02-01T23:59:59.000Z

    The baseline design for space nuclear power is a fission surface power (FSP) system: sodium-potassium (NaK) cooled, fast spectrum reactor with highly-enriched-uranium (HEU)-O2 fuel, stainless steel (SS) cladding, and beryllium reflectors with B4C control drums. Previous studies were performed to evaluate modeling capabilities and quantify uncertainties and biases associated with analysis methods and nuclear data. Comparison of Zero Power Plutonium Reactor (ZPPR)-20 benchmark experiments with the FSP design indicated that further reduction of the total design model uncertainty requires the reduction in uncertainties pertaining to beryllium and uranium cross-section data. Further comparison with three beryllium-reflected HEU-metal benchmark experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) concluded the requirement that experimental validation data have similar cross section sensitivities to those found in the FSP design. A series of critical experiments was performed at ORCEF in the 1960s to support the Medium Power Reactor Experiment (MPRE) space reactor design. The small, compact critical assembly (SCCA) experiments were graphite- or beryllium-reflected assemblies of SS-clad, HEU-O2 fuel on a vertical lift machine. All five configurations were evaluated as benchmarks. Two of the five configurations were beryllium reflected, and further evaluated using the sensitivity and uncertainty analysis capabilities of SCALE 6.1. Validation of the example FSP design model was successful in reducing the primary uncertainty constituent, the Be(n,n) reaction, from 0.28 %dk/k to 0.0004 %dk/k. Further assessment of additional reactor physics measurements performed on the SCCA experiments may serve to further validate FSP design and operation.

  3. Modularity in design of the MIT Pebble Bed Reactor

    E-Print Network [OSTI]

    Berte, Marc Vincent, 1977-

    2004-01-01T23:59:59.000Z

    The future of new nuclear power plant construction will depend in large part on the ability of designers to reduce capital, operations, and maintenance costs. One of the methods proposed, is to enhance the modularity of ...

  4. DESIGN OF A MICROCHANNEL BASED SOLAR RECEIVER/REACTOR FOR

    E-Print Network [OSTI]

    Apte, Sourabh V.

    for production of hydrogen. A cou- pled shape-constrained optimization and Monte-Carlo radiative heat transfer model is developed to design a receiver shape that can yield a desired heat flux distribution

  5. Conceptual design of a pressure tube light water reactor with variable moderator control

    SciTech Connect (OSTI)

    Rachamin, R.; Fridman, E. [Reactor Safety Div., Inst. of Resource Ecology, Helmholtz-Zentrum Dresden-Rossendorf, POB 51 01 19, 01314 Dresden (Germany); Galperin, A. [Dept. of Nuclear Engineering, Ben-Gurion Univ. of the Negev, POB 653, Beer Sheva 84105 (Israel)

    2012-07-01T23:59:59.000Z

    This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

  6. Design of Batch Tube Reactor 377 Applied Biochemistry and Biotechnology Vol. 9193, 2001

    E-Print Network [OSTI]

    California at Riverside, University of

    Design of Batch Tube Reactor 377 Applied Biochemistry and Biotechnology Vol. 91­93, 2001 Copyright unparalleled environmental, economic, and strategic benefits. However, low-cost, high-yield technologies for varying reaction con- ditions. In this article, heat transfer for batch tubes is analyzed to derive

  7. Design, Analysis and Optimization of the Power Conversion System for the Modular Pebble Bed Reactor System

    E-Print Network [OSTI]

    Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power

  8. Spring design for use in the core of a nuclear reactor

    DOE Patents [OSTI]

    Willard, Jr., H. James (Bethel Park, PA)

    1993-01-01T23:59:59.000Z

    A spring design particularly suitable for use in the core of a nuclear reactor includes one surface having a first material oriented in a longitudinal direction, and another surface having a second material oriented in a transverse direction. The respective surfaces exhibit different amounts of irraditation induced strain.

  9. Plasma engineering design of a compact reversed-field pinch reactor (CRFPR)

    SciTech Connect (OSTI)

    Bathke, C.G.; Embrechts, M.J.; Hagenson, R.L.; Krakowski, R.A.; Miller, R.L.

    1983-01-01T23:59:59.000Z

    The rationale for and the characteristics of the high-power-density Compact Reversed-Field Pinch Reactor (CRFPR) are discussed. Particular emphasis is given to key plasma engineering aspects of the conceptual design, including plasma operations, current drive, and impurity/ash control by means of pumped limiters or magnetic divertors. A brief description of the Fusion-Power-Core integration is given.

  10. Strategic planning for LEED certification

    E-Print Network [OSTI]

    Capuzzi, Angelo Michael

    2010-01-01T23:59:59.000Z

    Intel Corporation has recently implemented a "green building" policy, which states that Intel will design all new facilities to achieve a minimum LEED-Silver certification. LEED (Leadership in Energy and Environmental ...

  11. Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

    SciTech Connect (OSTI)

    Marcille, T. F.; Poston, D. I.; Kapernick, R. J. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Dixon, D. D. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Fischer, G. A. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Doherty, S. P. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Engineering, Trinity College, Hartford, CT 06106 (United States)

    2006-01-20T23:59:59.000Z

    In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

  12. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    SciTech Connect (OSTI)

    Ingersoll, D.T.

    2004-07-29T23:59:59.000Z

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the equivalent temperature of heat delivered to either the power conversion system or a hydrogen production plant. Using a comparative cost analysis, the construction costs per unit output are projected to be 50-55% of the costs for modular gas-cooled or sodium-cooled reactor systems. This is primarily a consequence of substantially larger power output and higher conversion efficiency for the AHTR. The AHTR has a number of unique technical challenges in meeting the NGNP requirements; however, it appears to offer advantages over high-temperature helium-cooled reactors and provides an alternative development path to achieve the NGNP requirements. Primary challenges include optimizing the core design for improved response to transients, designing an internal blanket to thermally protect the reactor vessel, and engineering solutions to high-temperature refueling and maintenance.

  13. Factors in reactor design for carbon dioxide capture with solid, regenerable sorbents

    SciTech Connect (OSTI)

    Hoffman, J.S.; Richards, G.A.; Pennline, H.W.; Fischer, Daniel (Mid-Atlantic Technology Research & Innovation Center, South Charleston, WV); Keller, George (Mid-Atlantic Technology Research & Innovation Center, South Charleston, WV)

    2008-06-01T23:59:59.000Z

    Fossil-fuel burning power plants, which produce a substantial amount of electric power within the United States, are point sources that can emit significant quantities of carbon dioxide (CO2). In a carbon sequestration scenario, the CO2 must first be captured from the point source, or flue gas, and then be permanently stored. Since the capture/separation step dominates the cost of sequestration, various capture/separation technologies are being investigated, and regenerable, solid sorbents are the basis for one promising technique for capturing CO2 from flue gas. The solid sorbent must be able to absorb the CO2 in the first step and then be regenerated by releasing the CO2 in the second step. Due to the low operating pressure of a conventional pulverized coal-fired combustor and its subsequent low partial pressure of CO2, it is envisioned that temperature swing absorption is applicable to the sorbent capture technology. Various CO2 capture sorbents are being examined in this research area, for example physical adsorbents as well as chemical absorbents. However, with respect to process development, various reactor configurations are presently being considered. The reactor designs range from stationary beds of sorbent to those systems where the sorbent is transported between the absorber and regenerator. Emphasis is placed on design implications of employing a regenerable solid sorbent system. Key sorbent parameters required for the sorbents have been identified, including the heat of adsorption, heat capacity of the solid, delta CO2 loading between the absorption and regeneration steps, and any role co-sorption of competitive gases, such as moisture, may play. Other sorbent properties, such as the effect of acid gases within the flue gas or the attrition of the sorbent, must be considered in the reactor design. These factors all impact the reactor design for a particular type of sorbent. For a generic sorbent, reactor designs have been formulated, including a stationary, isothermal reactor, a fluidized bed, and a moving bed. Through calculations, benefits and disadvantages of the designs have been outlined. The implication of the sorbent properties (and thus desired experimental information) on sorbent reactor design are described, and recommendations for operation of these types of capture systems are discussed.

  14. Design of irradiation rig for reactor testing of prototype bolometers for ITER

    SciTech Connect (OSTI)

    Gusarov, A.; Huysmans, S. [SCK.CEN Belgian Nucrear Research Center, 2400 Mol (Belgium); Meister, H. [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, D-85748 Garching b. Muenchen (Germany); Hodgson, E. [Euratom/CIEMAT Fusion Association, Avenida Complutense 22, 28040 Madrid (Spain)

    2011-07-01T23:59:59.000Z

    We describe the design of an experimental rig, which was developed to allow reactor testing at relevant conditions, i.e. vacuum and {approx}400 deg.C temperature, of prototype resistive bolometers, which will be used in ITER to acquire information on the radiated power distribution from the main plasma and in the diverter region. The main feature of the design is that the rig has no active temperature control. (authors)

  15. Environmental Certification

    E-Print Network [OSTI]

    Environmental Compliance Inspector Certification 2009 Candidate Handbook This booklet contains... ° Subject matter for the Environmental Compliance Inspector tests ° Education and experience requirements Contents Environmental Compliance Inspector 2009 Candidate Handbook This handbook contains information

  16. Options Study Documenting the Fast Reactor Fuels Innovative Design Activity

    SciTech Connect (OSTI)

    Jon Carmack; Kemal Pasamehmetoglu

    2010-07-01T23:59:59.000Z

    This document provides presentation and general analysis of innovative design concepts submitted to the FCRD Advanced Fuels Campaign by nine national laboratory teams as part of the Innovative Transmutation Fuels Concepts Call for Proposals issued on October 15, 2009 (Appendix A). Twenty one whitepapers were received and evaluated by an independent technical review committee.

  17. Secondary heat exchanger design and comparison for advanced high temperature reactor

    SciTech Connect (OSTI)

    Sabharwall, P. [Idaho National Laboratory, Idaho Falls, ID 83415-3860 (United States); Kim, E. S. [Seoul National Univ., P.O. Box 1625, Idaho Falls, ID 83415-3860 (United States); Siahpush, A.; McKellar, M.; Patterson, M. [Idaho National Laboratory, Idaho Falls, ID 83415-3860 (United States)

    2012-07-01T23:59:59.000Z

    Next generation nuclear reactors such as the advanced high temperature reactor (AHTR) are designed to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. This study considers two different types of heat exchangers - helical coiled heat exchanger and printed circuit heat exchanger - as possible options for the AHTR secondary heat exchangers with distributed load analysis and comparison. Comparison is provided for all different cases along with challenges and recommendations. (authors)

  18. A helium-cooled blanket design of the low aspect ratio reactor

    SciTech Connect (OSTI)

    Wong, C.P.; Baxi, C.B.; Reis, E.E. [General Atomics, San Diego, CA (United States); Cerbone, R.; Cheng, E.T. [TSI Research, Solana Beach, CA (United States)

    1998-03-01T23:59:59.000Z

    An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh.

  19. Computational Science and Engineering Certification for Nuclear, Plasma, and Radiological Engineering

    E-Print Network [OSTI]

    Gilbert, Matthew

    Computational Science and Engineering Certification for Nuclear, Plasma, and Radiological Engineering The Computational Science and Engineering certificate program is designed to provide NPRE a certificate in "Computational Science and Engineering", students must complete the required courses listed

  20. Overview of Fusion-Fission Hybrid Reactor Design Study in China

    SciTech Connect (OSTI)

    Huang Jinhua [Southwestern Institute of Physics (China); Feng Kaiming [Southwestern Institute of Physics (China); Deng Baiquan [Southwestern Institute of Physics (China); Deng, P.Zh. [Southwestern Institute of Physics (China); Zhang Guoshu [Southwestern Institute of Physics (China); Hu Gang [Southwestern Institute of Physics (China); He Kaihui [Southwestern Institute of Physics (China); Wu Yican [Institute of Plasma Physics (China); Qiu Lijian [Institute of Plasma Physics (China); Huang Qunying [Institute of Plasma Physics (China); Xiao Bingjia [Institute of Plasma Physics (China); Liu Xiaoping [Institute of Plasma Physics (China); Chen Yixue [Institute of Plasma Physics (China); Kong, M.H. [Institute of Plasma Physics (China)

    2002-07-15T23:59:59.000Z

    The motivation for developing fusion-fission hybrid reactors is discussed in the context of electricity power requirements by 2050 in China. A detailed conceptual design of the Fusion Experimental Breeder (FEB) was developed from 1986-1995. The FEB has a subignited tokamak fusion core with a major radius of 4.0 m, a fusion power of 145 MW, and a fusion energy gain Q of 3. Based on this, an engineering outline design study of the FEB, FEB-E, has been performed. This design study is a transition from conceptual to engineering design in this research. The main results beyond that given in the detailed conceptual design are included in this paper, namely, the design studies of the blanket, divertor, test blanket, and tritium and environment issues. In-depth analyses have been performed to support the design. Studies of related advanced concepts such as the waste transmutation blanket concept and the spherical tokamak core concept are also presented.

  1. Design of an Alternative Coolant Inlet Flow Configuaration for the Modular Helium Reactor

    SciTech Connect (OSTI)

    SM Mohsin Reza; E. A. Harvego; Matt Richards; Arkal Shenoy; Kenneth Lee Peddicord

    2006-06-01T23:59:59.000Z

    The coolant outlet temperature for the Modular Helium Reactor (MHR) was increased to improve the overall efficiency of nuclear hydrogen production using either thermochemical or high temperature electrolysis (HTE) processes. The inlet temperature was also increased to keep about the same _T across the reactor core. Thermal hydraulic analyses of the current MHR design were performed with these updated temperatures to determine the impact of these highter temperatures on pressure drops, coolant flow rates and temperature profiles within the vessel and core regions. Due to these increased operating temperatures, the overall efficiency of hydrogen production processes increases but the steady state reactor vessel temperature is found to be well above the ASME code limits for current vessel materials. Using the RELAP5-3D/ATHENA computer code, an alternative configuration for the MHR coolant inlet flow path was evaluated in an attempt to reduce the reactor vessel temperatures. The coolant inlet flow was shifted from channel boxes located in the annular region between the reactor core barrel and the inner wall of the reactor vessel to a flow path through the outer permanent reflector. Considering the available thickness of graphite in the permanent outer reflector, the total flow area, the number of coolant holes and the coolant-hole diameter were varied to optimize the pressure drop, the coolant inlet velocity and the percentage of graphite removed from the core. The resulting thermal hydraulic analyses of the optimized design showed that peak vessel and fuel temperatures were within acceptable limits for both steady-state and transient operating conditions.

  2. KEY DESIGN REQUIREMENTS FOR THE HIGH TEMPERATURE GAS-COOLED REACTOR NUCLEAR HEAT SUPPLY SYSTEM

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01T23:59:59.000Z

    Key requirements that affect the design of the high temperature gas-cooled reactor nuclear heat supply system (HTGR-NHSS) as the NGNP Project progresses through the design, licensing, construction and testing of the first of a kind HTGR based plant are summarized. These requirements derive from pre-conceptual design development completed to-date by HTGR Suppliers, collaboration with potential end users of the HTGR technology to identify energy needs, evaluation of integration of the HTGR technology with industrial processes and recommendations of the NGNP Project Senior Advisory Group.

  3. Design and operation of a rotating drum radio frequency plasma reactor for the modification of free nanoparticles

    SciTech Connect (OSTI)

    Shearer, Jeffrey C.; Fisher, Ellen R. [Department of Chemistry, Colorado State University, Fort Collins, Colorado 80523-1872 (United States)

    2013-06-15T23:59:59.000Z

    A rotating drum rf plasma reactor was designed to functionalize the surface of nanoparticles and other unusually shaped substrates through plasma polymerization and surface modification. This proof-of-concept reactor design utilizes plasma polymerized allyl alcohol to add OH functionality to Fe{sub 2}O{sub 3} nanoparticles. The reactor design is adaptable to current plasma hardware, eliminating the need for an independent reactor setup. Plasma polymerization performed on Si wafers, Fe{sub 2}O{sub 3} nanoparticles supported on Si wafers, and freely rotating Fe{sub 2}O{sub 3} nanoparticles demonstrated the utility of the reactor for a multitude of processes. X-ray photoelectron spectroscopy and Fourier transform infrared spectroscopy were used to characterize the surface of the substrates prior to and after plasma deposition, and scanning electron microscopy was used to verify that no extensive change in the size or shape of the nanoparticles occurred because of the rotating motion of the reactor. The reactor design was also extended to a non-depositing NH{sub 3} plasma modification system to demonstrate the reactor design is effective for multiple plasma processes.

  4. Conceptual design characteristics of a denatured molten-salt reactor with once-through fueling

    SciTech Connect (OSTI)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E.; Rhoades, W.A.

    1980-07-01T23:59:59.000Z

    A study was made to examine the conceptual feasibility of a molten-salt power reactor fueled with denatured /sup 235/U and operated with a minimum of chemical processing. Because such a reactor would not have a positive breeding gain, reductions in the fuel conversion ratio were allowed in the design to achieve other potentially favorable characteristics for the reactor. A conceptual core design was developed in which the power density was low enough to allow a 30-year life expectancy of the moderator graphite with a fluence limit of 3 x 10/sup 26/ neutrons/m/sup 2/ (E > 50 keV). This reactor could be made critical with about 3450 kg of 20% enriched /sup 235/U and operated for 30 years with routine additions of denatured /sup 235/U and no chemical processing for removal of fission products. A review of the chemical considerations assoicated with the conceptual fuel cycle indicates that no substantial difficulties would be expected if the soluble fission products and higher actinides were allowed to remain in the fuel salt for the life of the plant.

  5. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    SciTech Connect (OSTI)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01T23:59:59.000Z

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost.

  6. An investigation of RVACS (reactor vessel auxiliary cooling system) design improvements

    SciTech Connect (OSTI)

    Tzanos, C.P.; Tessier, J.H.; Pedersen, D.R. (Argonne National Laboratory, IL (USA))

    1989-11-01T23:59:59.000Z

    One of the main safety features of the current liquid-metal reactor (LMR) designs is the utilization of decay heat removal systems that remove heat by natural convection. In the reactor vessel auxiliary cooling system (RVACS), decay heat is removed by naturally circulating air in the gap between the guard vessel and a baffle wall surrounding the guard vessel. The objective of this work was to determine the impact of a number of design parameters on the performance of the RVACS of a pool LMR. These parameters were (a) the stack height, (b) the size of the airflow gap, (c) the system pressure loss, (d) fins on the guard vessel or the baffle wall, and (e) roughness (in the form of repeated ribs) on the airflow channel walls. Reactor designs ranging from 400 to 3,500 MW(thermal) were considered. From the RVACS design parameters considered in this analysis, an optimized ribbed configuration gave the best improvement in RVACS performance. For a 3,500-MW(thermal) LMR, the peak sodium and cladding temperatures were reduced by 52 K.

  7. Evaluation of a Business Case for Safeguards by Design in Nuclear Power Reactors

    SciTech Connect (OSTI)

    Wood, Thomas W.; Seward, Amy M.; Lewis, Valerie A.; Gitau, Ernest TN; Zentner, Michael D.

    2012-12-01T23:59:59.000Z

    Safeguards by Design (SbD) is a well-known paradigm for consideration and incorporation of safeguards approaches and associated design features early in the nuclear facility development process. This paradigm has been developed as part of the Next Generation Safeguards Initiative (NGSI), and has been accepted as beneficial in many discussions and papers on NGSI or specific technologies under development within NGSI. The Office of Nuclear Safeguards and Security funded the Pacific Northwest National Laboratory to examine the business case justification of SbD for nuclear power reactors. Ultimately, the implementation of SbD will rely on the designers of nuclear facilities. Therefore, it is important to assess the incentives which will lead designers to adopt SbD as a standard practice for nuclear facility design. This report details the extent to which designers will have compelling economic incentives to adopt SbD.

  8. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    L. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,R. J. Neuhold, Introductury Nuclear Reactor Dynamics. ANSL. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,

  9. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    SciTech Connect (OSTI)

    Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

    1981-01-01T23:59:59.000Z

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

  10. Development and evaluation of two reactor designs for desulfurization of Texas lignites

    E-Print Network [OSTI]

    Merritt, Stanley Duane

    1991-01-01T23:59:59.000Z

    exhibited can be given at this time, but this behavior may be indicative of transformations of inorganic matter, changes in the forms of sulfur present in the lignite, and the overall composition. The results of this test series show a need for further...DEVELOPMENT AND EVALUATION OF TWO REACTOR DESIGNS FOR DESULFURIZATION OF TEXAS LIGNITES A Thesis by STANLEY DUANE MERRITT Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements...

  11. Seismicity and seismic response of the Soviet-designed VVER (Water-cooled, Water moderated Energy Reactor) reactor plants

    SciTech Connect (OSTI)

    Ma, D.C.; Gvildys, J.; Wang, C.Y.; Spencer, B.W.; Sienicki, J.J.; Seidensticker, R.W.; Purvis, E.E. III

    1989-01-01T23:59:59.000Z

    On March 4, 1977, a strong earthquake occurred at Vrancea, Romania, about 350 km from the Kozloduy plant in Bulgaria. Subsequent to this event, construction of the unit 2 of the Armenia plant was delayed over two years while seismic features were added. On December 7, 1988, another strong earthquake struck northwest Armenia about 90 km north of the Armenia plant. Extensive damage of residential and industrial facilities occurred in the vicinity of the epicenter. The earthquake did not damage the Armenia plant. Following this event, the Soviet government announced that the plant would be shutdown permanently by March 18, 1989, and the station converted to a fossil-fired plant. This paper presents the results of the seismic analyses of the Soviet-designed VVER (Water-cooled, Water moderated Energy Reactor) plants. Also presented is the information concerning seismicity in the regions where VVERs are located and information on seismic design of VVERs. The reference units are the VVER-440 model V230 (similar to the two units of the Armenia plant) and the VVER-1000 model V320 units at Kozloduy in Bulgaria. This document provides an initial basis for understanding the seismicity and seismic response of VVERs under seismic events. 1 ref., 9 figs., 3 tabs.

  12. Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters

    SciTech Connect (OSTI)

    Frank Darmann; Robert Lombaerde; Franco Moriconi; Albert Nelson

    2011-10-31T23:59:59.000Z

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with â??warm boreâ?ť diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged â??spiderâ?ť design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project â??Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limitersâ?ť was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZPâ??s product development program, the amount of HTS wire employed per FCL and its cost as a percentage of the total FCL product content had not dropped substantially from an unsustainable level of more than 50% of the total cost of the FCL, nor had the availability increased (today the availability of 2G wire for commercial applications outside of specific partnerships with the leading 2G wire manufacturers is extremely limited). ZP had projected a very significant commercial potential for FCLs with higher performance and lower costs compared to the initial models built with 1G wire, which would come about from the widespread availability of low-cost, high-performance 2G HTS wire. The potential for 2G wires at greatly reduced performance-based prices compared to 1G HTS conductor held out the potential for the commercial production of FCLs at price and performance levels attractive to the utility industry. However, the price of HTS wire did not drop as expected and today the available quantities of 2G wire are limited, and the price is higher than the currently available supplies of 1G wire. The commercial option for ZP to provide a reliable and reasonably priced FCL to the utility industry is to employ conventional resistive conductor DC electromagnets to bias the FCL. Since the premise of the original funding was to stimulate the HTS wire industry and ZP concluded that copper-based magnets were more economical for the foreseeable future, DOE and ZP decided to mutually terminate the project.

  13. A Study of Fast Reactor Fuel Transmutation in a Candidate Dispersion Fuel Design

    SciTech Connect (OSTI)

    Mark DeHart; Hongbin Zhang; Eric Shaber; Matthew Jesse

    2010-11-01T23:59:59.000Z

    Dispersion fuels represent a significant departure from typical ceramic fuels to address swelling and radiation damage in high burnup fuel. Such fuels use a manufacturing process in which fuel particles are encapsulated within a non-fuel matrix. Dispersion fuels have been studied since 1997 as part of an international effort to develop and test very high density fuel types for the Reduced Enrichment for Research and Test Reactors (RERTR) program.[1] The Idaho National Laboratory is performing research in the development of an innovative dispersion fuel concept that will meet the challenges of transuranic (TRU) transmutation by providing an integral fission gas plenum within the fuel itself, to eliminate the swelling that accompanies the irradiation of TRU. In this process, a metal TRU vector produced in a separations process is atomized into solid microspheres. The dispersion fuel process overcoats the microspheres with a mixture of resin and hollow carbon microspheres to create a TRUC. The foam may then be heated and mixed with a metal power (e.g., Zr, Ti, or Si) and resin to form a matrix metal carbide, that may be compacted and extruded into fuel elements. In this paper, we perform reactor physics calculations for a core loaded with the conceptual fuel design. We will assume a “typical” TRU vector and a reference matrix density. We will employ a fuel and core design based on the Advanced Burner Test Reactor (ABTR) design.[2] Using the CSAS6 and TRITON modules of the SCALE system [3] for preliminary scoping studies, we will demonstrate the feasibility of reactor operations. This paper will describe the results of these analyses.

  14. Design modification for the modular helium reactor for higher temperature operation and reliability studies for nuclear hydrogen production processes

    E-Print Network [OSTI]

    Reza, S.M. Mohsin

    2009-05-15T23:59:59.000Z

    Design options have been evaluated for the Modular Helium Reactor (MHR) for higher temperature operation. An alternative configuration for the MHR coolant inlet flow path is developed to reduce the peak vessel temperature (PVT). The coolant inlet...

  15. Investigation and design of a secure, transportable fluoride-salt-cooled high-temperature reactor (TFHR) for isolated locations

    E-Print Network [OSTI]

    Macdonald, Ruaridh (Ruaridh R.)

    2014-01-01T23:59:59.000Z

    In this work we describe a preliminary design for a transportable fluoride salt cooled high temperature reactor (TFHR) intended for use as a variable output heat and electricity source for off-grid locations. The goals of ...

  16. Analysis of Reference Design for Nuclear-Assisted Hydrogen Production at 750°C Reactor Outlet Temperature

    SciTech Connect (OSTI)

    Michael G. McKellar; Edwin A. Harvego

    2010-05-01T23:59:59.000Z

    The use of High Temperature Electrolysis (HTE) for the efficient production of hydrogen without the greenhouse gas emissions associated with conventional fossil-fuel hydrogen production techniques has been under investigation at the Idaho National Engineering Laboratory (INL) for the last several years. The activities at the INL have included the development, testing and analysis of large numbers of solid oxide electrolysis cells, and the analyses of potential plant designs for large scale production of hydrogen using a high-temperature gas-cooled reactor (HTGR) to provide the process heat and electricity to drive the electrolysis process. The results of this research led to the selection in 2009 of HTE as the preferred concept in the U.S. Department of Energy (DOE) hydrogen technology down-selection process. However, the down-selection process, along with continued technical assessments at the INL, has resulted in a number of proposed modifications and refinements to improve the original INL reference HTE design. These modifications include changes in plant configuration, operating conditions and individual component designs. This report describes the resulting new INL reference design coupled to two alternative HTGR power conversion systems, a Steam Rankine Cycle and a Combined Cycle (a Helium Brayton Cycle with a Steam Rankine Bottoming Cycle). Results of system analyses performed to optimize the design and to determine required plant performance and operating conditions when coupled to the two different power cycles are also presented. A 600 MWt high temperature gas reactor coupled with a Rankine steam power cycle at a thermal efficiency of 44.4% can produce 1.85 kg/s of hydrogen and 14.6 kg/s of oxygen. The same capacity reactor coupled with a combined cycle at a thermal efficiency of 42.5% can produce 1.78 kg/s of hydrogen and 14.0 kg/s of oxygen.

  17. Light Water Breeder Reactor fuel rod design and performance characteristics (LWBR Development Program)

    SciTech Connect (OSTI)

    Campbell, W.R.; Giovengo, J.F.

    1987-10-01T23:59:59.000Z

    Light Water Breeder Reactor (LWBR) fuel rods were designed to provide a reliable fuel system utilizing thorium/uranium-233 mixed-oxide fuel while simultaneously minimizing structural material to enhance fuel breeding. The fuel system was designed to be capable of operating successfully under both load follow and base load conditions. The breeding objective required thin-walled, low hafnium content Zircaloy cladding, tightly spaced fuel rods with a minimum number of support grid levels, and movable fuel rod bundles to supplant control rods. Specific fuel rod design considerations and their effects on performance capability are described. Successful completion of power operations to over 160 percent of design lifetime including over 200 daily load follow cycles has proven the performance capability of the fuel system. 68 refs., 19 figs., 44 tabs.

  18. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    SciTech Connect (OSTI)

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01T23:59:59.000Z

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  19. Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core

    SciTech Connect (OSTI)

    Sterbentz, James W

    2007-05-01T23:59:59.000Z

    A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.

  20. Design of a Gas Test Loop Facility for the Advanced Test Reactor

    SciTech Connect (OSTI)

    C. A. Wemple

    2005-09-01T23:59:59.000Z

    The Office of Nuclear Energy within the U.S. Department of Energy (DOE-NE) has identified the need for irradiation testing of nuclear fuels and materials, primarily in support of the Generation IV (Gen-IV) and Advanced Fuel Cycle Initiative (AFCI) programs. These fuel development programs require a unique environment to test and qualify potential reactor fuel forms. This environment should combine a high fast neutron flux with a hard neutron spectrum and high irradiation temperature. An effort is presently underway at the Idaho National Laboratory (INL) to modify a large flux trap in the Advanced Test Reactor (ATR) to accommodate such a test facility [1,2]. The Gas Test Loop (GTL) Project Conceptual Design was initiated to determine basic feasibility of designing, constructing, and installing in a host irradiation facility, an experimental vehicle that can replicate with reasonable fidelity the fast-flux test environment needed for fuels and materials irradiation testing for advanced reactor concepts. Such a capability will be needed if programs such as the AFCI, Gen-IV, the Next Generation Nuclear Plant (NGNP), and space nuclear propulsion are to meet development objectives and schedules. These programs are beginning some irradiations now, but many call for fast flux testing within this decade.

  1. Integrating Safety, Operations, Security, and Safeguards (ISOSS) into the design of small modular reactors : a handbook.

    SciTech Connect (OSTI)

    Middleton, Bobby D.; Mendez, Carmen Margarita [Sociotecnia Solutions] [Sociotecnia Solutions

    2013-10-01T23:59:59.000Z

    The existing regulatory environment for nuclear reactors impacts both the facility design and the cost of operations once the facility is built. Delaying the consideration of regulatory requirements until late in the facility design - or worse, until after construction has begun - can result in costly retrofitting as well as increased operational costs to fulfill safety, security, safeguards, and emergency readiness requirements. Considering the scale and scope, as well as the latest design trends in the next generation of nuclear facilities, there is an opportunity to evaluate the regulatory requirements and optimize the design process for Small Modular Reactors (SMRs), as compared to current Light Water Reactors (LWRs). To this end, Sandia has embarked on an initiative to evaluate the interactions of regulations and operations as an approach to optimizing the design of SMR facilities, supporting operational efficiencies, as well as regulatory requirements. The early stages of this initiative consider two focus areas. The first focus area, reported by LaChance, et al. (2007), identifies the regulatory requirements established for the current fleet of LWR facilities regarding Safety, Security, Operations, Safeguards, and Emergency Planning, and evaluates the technical bases for these requirements. The second focus area, developed in this report, documents the foundations for an innovative approach that supports a design framework for SMR facilities that incorporates the regulatory environment, as well as the continued operation of the facility, into the early design stages, eliminating the need for costly retrofitting and additional operating personnel to fulfill regulatory requirements. The work considers a technique known as Integrated Safety, Operations, Security and Safeguards (ISOSS) (Darby, et al., 2007). In coordination with the best practices of industrial operations, the goal of this effort is to develop a design framework that outlines how ISOSS requirements can be incorporated into the pre-conceptual through early facility design stages, seeking a cost-effective design that meets both operational efficiencies and the regulatory environment. The larger scope of the project, i.e., in future stages, includes the identification of potentially conflicting requirements identified by the ISOSS framework, including an analysis of how regulatory requirements may be changed to account for the intrinsic features of SMRs.

  2. Design and analysis of megawatt-class heat-pipe reactor concepts

    SciTech Connect (OSTI)

    Poston, D.; Kapernick, R. [Los Alamos National Laboratory, MS C921, Los Alamos, NM 87545 (United States)

    2012-07-01T23:59:59.000Z

    There is growing interest in finding an alternative to diesel-powered systems at locations removed from a reliable electrical grid. One promising option is a 1- to 10-MW mobile reactor system, that could provide robust, self-contained, and long-term ({>=} 5 years) power in any environment. The reactor and required infrastructure could be transported to any location within one or a few standard transport containers. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than 'traditional' reactors that rely on pumped coolant through the core. This paper examines a heat pipe reactor that is fabricated and shipped as six identical core segments. Each core segment includes a heat-pipe-to-gas heat exchanger that is coupled to the condenser end of the heat pipes. The reference power conversion system is a CO{sub 2}-Brayton system. The segments by themselves are deeply subcritical during transport, and they would be locked into an operating configuration (with control inserted) at the final destination. Two design options are considered: a near-term option and an advanced option. The near-term option is a 5-MWt concept that uses uranium-dioxide fuel, a stainless-steel structure, and potassium as the heat-pipe working fluid. The advanced option is a 15-MWt concept that uses uranium-nitride fuel, a molybdenum/TZM structure, and sodium as the heat-pipe working fluid. The materials used in the advanced option allow for higher temperatures and power densities, and enhanced power throughput in the heat pipes. Higher powers can be obtained from both concepts by increasing the core size and the number of heat pipes. (authors)

  3. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    SciTech Connect (OSTI)

    Adams, S.R.

    1985-10-01T23:59:59.000Z

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  4. The Treatment of Renewable Energy Certificates, Emissions Allowances, and Green Power Programs in State Renewables Portfolio Standards

    E-Print Network [OSTI]

    Holt, Edward A.; Wiser, Ryan H.

    2007-01-01T23:59:59.000Z

    Design Guide for Renewable Energy Certificate Tracking2005. “Emerging Markets for Renewable Energy Certificates:Golden, CO: National Renewable Energy Laboratory. http://

  5. The Neutronics Design and Analysis of a 200-MW(electric) Simplified Boiling Water Reactor Core

    SciTech Connect (OSTI)

    Tinkler, Daniel R.; Downar, Thomas J. [Purdue University (United States)

    2003-06-15T23:59:59.000Z

    A 200-MW(electric) simplified boiling water reactor (SBWR) was designed and analyzed under sponsorship of the U.S. Department of Energy Nuclear Energy Research Initiative program. The compact size of a 200-MW(electric) reactor makes it attractive for countries with a less well developed engineering infrastructure, as well as for developed countries seeking to tailor generation capacity more closely to the growth of their electricity demand. The 200-MW(electric) core design reported here is based on the 600-MW(electric) General Electric SBWR core, which was first analyzed in the work performed here in order to qualify the computer codes used in the analysis. Cross sections for the 8 x 8 fuel assembly design were generated with the HELIOS lattice physics code, and core simulation was performed with the U.S. Nuclear Regulatory Commission codes RELAP5/PARCS. In order to predict the critical heat flux, the Hench-Gillis correlation was implemented in the RELAP5 code. An equilibrium cycle was designed for the 200-MW(electric) core, which provided a cycle length of more than 2 yr and satisfied the minimum critical power ratio throughout the core life.

  6. Design of a Simplified Closed Brayton Cycle for a Space Reactor Application

    SciTech Connect (OSTI)

    Guimaraes, Lamartine N. F. [Institute for Advanced Studies-IEAv, Rodovia dos Tamoios, km 5.5, Putim, 12228-001 Sao Jose dos Campos, SP (Brazil); Faculdade de Tecnologia Sao Francisco Jacarei, SP, Brazil 55-12-3947-5474 (Brazil); Camillo, Giannino Ponchio [Institute for Advanced Studies-IEAv, Rodovia dos Tamoios, km 5.5, Putim, 12228-001 Sao Jose dos Campos, SP (Brazil); Placco, Guilherme Moreira [Faculdade de Tecnologia Sao Francisco Jacarei, SP, Brazil 55-12-3947-5474 (Brazil)

    2009-03-16T23:59:59.000Z

    The Nuclear Energy Division (ENU) of the Institute for Advanced Studies (IEAv) has started a preliminary design study for a Closed Brayton Cycle Loop (CBCL) aimed at a space reactor application. The main objectives of the study are: 1) to establish a starting concept for the CBCL components specifications, and 2) to build a demonstrative simulator of CBCL. This preliminary design study is been developed around the NOELLE 60290 turbo machine. The actual nuclear reactor study is being conducted independently. Because of that, a conventional heat source is being used for the CBCL, in this preliminary design phase. This paper describes details of the CBCL mechanical design and the steady state simulator of the CBCL operating with NOELLE 60290 turbo machine. In principle, several gases are being considered as working fluid, as for instance: air, helium, nitrogen, CO2 and gas mixtures such as helium and xenon. However, for this first application pure helium will be used as working fluid. Simplified models of heat and mass transfer were developed to simulate thermal components. A new graphical interface was developed for the simulator to display the thermal process variables in steady state and to keep track of the modifications being implemented at the NOELLE 60290 turbo machine in order to build the CBCL. A set of new results are being produced. These new results help to establish the hot and cold source geometry allowing for price estimating costs for building the actual device. These fresh new results will be presented and discussed.

  7. Environmental Certificate Program

    E-Print Network [OSTI]

    Rose, Michael R.

    Environmental Management Certificate Program Accelerate Your Career Environmentaland Facilities of excellence. Environmental Management Certificate Program Compliance with regulatory requirements, remediation Irvine Extension's Certificate Program in Environmental Manage- ment prepares professionals at every

  8. Certificate Industrial and Systems Engineering

    E-Print Network [OSTI]

    Su, Xiao

    Six Sigma Certificate Industrial and Systems Engineering San José State University September, 2008 #12;1 Lean Enterprise and Six Sigma Lean Enterprise about transforming the old mass production-to-cradle design, incorporating design for manufacturability, reproducibility, product lifecycle, etc. Six Sigma

  9. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    SciTech Connect (OSTI)

    Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01T23:59:59.000Z

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  10. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    SciTech Connect (OSTI)

    M. Ishii; S. T. Revankar; T. Downar; Y. Xu, H. J. Yoon; D. Tinkler; U. S. Rohatgi

    2003-06-16T23:59:59.000Z

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral system scaling analysis, design parameters were obtained and designs of the compact modular 200 MWe SBWR and the full size 1200 MWe SBWR were developed. These reactors are provided with passive safety systems. A new passive vacuum breaker check valve was designed to replace the mechanical vacuum beaker check valve. The new vacuum breaker check valve was based on a hydrostatic head, and was fail safe. The performance of this new valve was evaluated both by the thermal-hydraulic code RELAP5 and by the experiments in a scaled SBWR facility, PUMA. In the core neutronic design a core depletion model was implemented to PARCS code. A lattice design for the SBWR fuel assemblies was performed. Design improvements were made to the neutronics/thermal-hydraulics models of SBWR-200 and SBWR-1200, and design analyses of these reactors were performed. The design base accident analysis and evaluation of all the passive safety systems were completed as scheduled in tasks 4 and 5. Initial conditions for the small break loss of coolant accidents (LOCA) and large break LOCA using REALP5 code were obtained. Small and large break LOCA tests were performed and the data was analyzed. An anticipated transient with scram was simulated using the RELAP5 code for SBWR-200. The transient considered was an accidental closure of the main steam isolation valve (MSIV), which was considered to be the most significant transient. The evaluation of the RELAP5 code against experimental data for SBWR-1200 was completed. In task 6, the instability analysis for the three SBWR designs (SBWR-1200, SBWR-600 and SBWR-200) were simulated for start-up transients and the results were similar. Neither the geysering instability, nor the loop type instability was predicted by RAMONA-4B in the startup simulation following the recommended procedure by GE. The density wave oscillation was not observed at all because the power level used in the simulation was not high enough. A study was made of the potential instabilities by imposing an unrealistically high power ramp in a short time period, as suggested by GE. RAMON

  11. Use of freeze-casting in advanced burner reactor fuel design

    SciTech Connect (OSTI)

    Lang, A. L.; Yablinsky, C. A.; Allen, T. R. [Dept. of Engineering Physics, Univ. of Wisconsin Madison, 1500 Engineering Drive, Madison, WI 53711 (United States); Burger, J.; Hunger, P. M.; Wegst, U. G. K. [Thayer School of Engineering, Dartmouth College, 8000 Cummings Hall, Hanover, NH 03755 (United States)

    2012-07-01T23:59:59.000Z

    This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models. Preliminary results show that criticality is achievable with freeze-cast fuel pins despite the significant amount of inert fuel matrix. Freeze casting is a promising method to achieve very precise fuel placement within fuel pins. (authors)

  12. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    to large conventional sodium fast reactors (SFR). TerraPowerincrease for a typical sodium fast reactor fuel rod geometryof the new Russian sodium fast reactor BN-800 [111]. The

  13. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    AND-BURN REACTOR PHYSICS wave burnup principle. The CANDLEand physical principle Breed-and-burn reactors (B&B) areBURN REACTOR PHYSICS The FIMA burnup unit - principles and

  14. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    1] B. Farrar et. al. , Fast reactor decay heat removal:CA [2] B. Farrar et. al. , Fast reactor decay heat removal:They are 1) gas cooled fast reactors (GFR), 2) very high

  15. Cogeneration of Electricity and Potable Water Using The International Reactor Innovative And Secure (IRIS) Design

    SciTech Connect (OSTI)

    Ingersoll, D.T.; Binder, J.L.; Kostin, V.I.; Panov, Y.K.; Polunichev, V.; Ricotti, M.E.; Conti, D.; Alonso, G.

    2004-10-06T23:59:59.000Z

    The worldwide demand for potable water has been steadily growing and is projected to accelerate, driven by a continued population growth and industrialization of emerging countries. This growth is reflected in a recent market survey by the World Resources Institute, which shows a doubling in the installed capacity of seawater desalination plants every ten years. The production of desalinated water is energy intensive, requiring approximately 3-6 kWh/m3 of produced desalted water. At current U.S. water use rates, a dedicated 1000 MW power plant for every one million people would be required to meet our water needs with desalted water. Nuclear energy plants are attractive for large scale desalination application. The thermal energy produced in a nuclear plant can provide both electricity and desalted water without the production of greenhouse gases. A particularly attractive option for nuclear desalination is to couple a desalination plant with an advanced, modular, passively safe reactor design. The use of small-to-medium sized nuclear power plants allows for countries with smaller electrical grid needs and infrastructure to add new electrical and water capacity in more appropriate increments and allows countries to consider siting plants at a broader number of distributed locations. To meet these needs, a modified version of the International Reactor Innovative and Secure (IRIS) nuclear power plant design has been developed for the cogeneration of electricity and desalted water. The modular, passively safe features of IRIS make it especially well adapted for this application. Furthermore, several design features of the IRIS reactor will ensure a safe and reliable source of energy and water even for countries with limited nuclear power experience and infrastructure. The IRIS-D design utilizes low-quality steam extracted from the low-pressure turbine to boil seawater in a multi-effect distillation desalination plant. The desalination plant is based on the horizontal tube film evaporation design used successfully with the BN-350 nuclear plant in Aktau, Kazakhstan. Parametric studies have been performed to optimize the balance of plant design. Also, an economic analysis has been performed, which shows that IRIS-D should be able to provide electricity and clean water at highly competitive costs.

  16. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    systems for the Gas Cooled Fast Reactor (GCFR) includes theThey are 1) gas cooled fast reactors (GFR), 2) very high

  17. Certificate in Geospatial Technology The School of Earth, Environment, and Society offers a graduate certificate in Geospatial

    E-Print Network [OSTI]

    Moore, Paul A.

    Certificate in Geospatial Technology The School of Earth, Environment, and Society offers a graduate certificate in Geospatial Technology. The purpose of the certificate program is to provide graduate students with training in geospatial theory, research design, and its application in fields

  18. Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

    SciTech Connect (OSTI)

    Bruce G. Schnitzler

    2012-01-01T23:59:59.000Z

    Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.

  19. 147 SDSU Curriculum Guide 2010 Certificate Programs

    E-Print Network [OSTI]

    Ponce, V. Miguel

    147 SDSU Curriculum Guide 2010 Certificate Programs Types of Certificate Programs General Guidelines for All Certificate Programs Academic Certificate Programs Professional Certificate programs Cosponsored Certificate Programs Other Certificates #12;148 SDSU Curriculum Guide 2010 University Guidelines

  20. US ITER (International Thermonuclear Experimental Reactor) shield and blanket design activities

    SciTech Connect (OSTI)

    Baker, C.C.

    1988-08-01T23:59:59.000Z

    This paper summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. Primary tasks carried out during the past year include design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components, and issues regarding structural materials for an ITER device. The blanket concepts considered are the aqueous/Li salt solution, a water-cooled, solid breeder blanket, a helium-cooled, solid-breeder blanket, a blanket cooled by helium containing lithium-bearing particulates, and a blanket concept based on breeding tritium from He/sup 3/. 1 ref., 2 tabs.

  1. The role of risk management in the design of diagnostics for fusion reactors

    SciTech Connect (OSTI)

    Ingesson, L. C. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Collaboration: F4E Diagnostic Project Team

    2014-08-21T23:59:59.000Z

    A project-oriented approach is beneficial for the selection and design of viable diagnostics for fusion reactors because of the associated complex physical and organizational environment. The project-oriented approach includes rigorous risk management. The nature and impact of risks related to technical, organizational and commercial aspects in relation to the development of ITER diagnostics under EU responsibility are analyzed. The majority of risks are related to organizational aspects and technical feasibility issues. The experience with ITER is extrapolated to DEMO and beyond. It should not be taken for granted that technical solutions will be found, while a risk analysis of various diagnostic techniques with quantitative assessments undertaken early in the design of DEMO would be beneficial.

  2. Atmospheric Pressure Reactor System | EMSL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Atmospheric Pressure Reactor System Atmospheric Pressure Reactor System The atmospheric pressure reactor system is designed for testing the efficiency of various catalysts for the...

  3. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    SciTech Connect (OSTI)

    Bruce G. Schnitzler; Stanley K. Borowski

    2012-07-01T23:59:59.000Z

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.

  4. Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    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.

  7. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    breeder reactors typically operate with an inner core of high fissile content surrounded by breeding blankets

  8. Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01T23:59:59.000Z

    An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

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

    SciTech Connect (OSTI)

    Godfroy, Thomas J.; Bragg-Sitton, Shannon M. [NASA Marshall Space Flight Center, TD40, Huntsville, Alabama, 35812 (United States); University of Michgan, Dept. of Nuclear Engineering and Radiological Sciences, Ann Arbor MI 48109 (United States); Kapernick, Richard J. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2004-02-04T23:59:59.000Z

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

  10. Safeguards-by-Design: Guidance for High Temperature Gas Reactors (HTGRs) With Pebble Fuel

    SciTech Connect (OSTI)

    Philip Casey Durst; Mark Schanfein

    2012-08-01T23:59:59.000Z

    The following is a guidance document from a series prepared for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), under the Next Generation Safeguards Initiative (NGSI), to assist facility designers and operators in implementing international Safeguards-by-Design (SBD). SBD has two main objectives: (1) to avoid costly and time consuming redesign work or retrofits of new nuclear fuel cycle facilities and (2) to make the implementation of international safeguards more effective and efficient at such facilities. In the long term, the attainment of these goals would save industry and the International Atomic Energy Agency (IAEA) time, money, and resources and be mutually beneficial. This particular safeguards guidance document focuses on pebble fuel high temperature gas reactors (HTGR). The purpose of the IAEA safeguards system is to provide credible assurance to the international community that nuclear material and other specified items are not diverted from peaceful nuclear uses. The safeguards system consists of the IAEA’s statutory authority to establish safeguards; safeguards rights and obligations in safeguards agreements and additional protocols; and technical measures implemented pursuant to those agreements. Of foremost importance is the international safeguards agreement between the country and the IAEA, concluded pursuant to the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). According to a 1992 IAEA Board of Governors decision, countries must: notify the IAEA of a decision to construct a new nuclear facility as soon as such decision is taken; provide design information on such facilities as the designs develop; and provide detailed design information based on construction plans at least 180 days prior to the start of construction, and on "as-built" designs at least 180 days before the first receipt of nuclear material. Ultimately, the design information will be captured in an IAEA Design Information Questionnaire (DIQ), prepared by the facility operator, typically with the support of the facility designer. The IAEA will verify design information over the life of the project. This design information is an important IAEA safeguards tool. Since the main interlocutor with the IAEA in each country is the State Regulatory Authority/SSAC (or Regional Regulatory Authority, e.g. EURATOM), the responsibility for conveying this design information to the IAEA falls to the State Regulatory Authority/SSAC.

  11. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    SciTech Connect (OSTI)

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01T23:59:59.000Z

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  12. Nuclear design of small-sized high temperature gas-cooled reactor for developing countries

    SciTech Connect (OSTI)

    Goto, M.; Seki, Y.; Inaba, Y.; Ohashi, H.; Sato, H.; Fukaya, Y.; Tachibana, Y. [Japan Atomic Energy Agency, 4002, Oarai-machi, Higashi Ibaraki-gun, Ibaraki-ken 311-1394 (Japan)

    2012-07-01T23:59:59.000Z

    Japan Atomic Energy Agency (JAEA) has started a conceptual design of a small-sized HTGR with 50 MW thermal power (HTR50S), which is a first-of-a-kind commercial or demonstration plant of a small-sized HTGR to be deployed in developing countries such as Kazakhstan in the 2020's. The nuclear design of the HTR50S is performed by upgrading the proven technology of the High Temperature Engineering Test Reactor (HTTR) to reduce the cost for the construction. In the HTTR design, twelve kinds of fuel enrichment was used to optimize the power distribution, which is required to make the maximum fuel temperature below the thermal limitation during the burn-up period. However, manufacture of many kinds of fuel enrichment causes increase of the construction cost. To solve this problem, the present study challenges the nuclear design by reducing the number of fuel enrichment to as few as possible. The nuclear calculations were performed with SRAC code system whose validity was proven by the HTTR burn-up data. The calculation results suggested that the optimization of the power distribution was reasonably achieved and the maximum fuel temperature was kept below the limitation by using three kinds of fuel enrichment. (authors)

  13. 1 5-9.2002 . We report the core and reflector design parameters of reactor PIK needed for neutronic calculations.

    E-Print Network [OSTI]

    Titov, Anatoly

    PIK, needed for the reactor neutronic calculation. All up-to-date design modifications are taken÷òåíû ïîñëåäíèå èçìåíåíèÿ â êîíñòðóêöèè. Abstract We report the core and reflector design parameters of reactor PIK needed for neutronic calculations. The recent changes are taken into account. Ïðåïðèíò ¹2472, 12

  14. Preliminary core design studies for the advanced burner reactor over a wide range of conversion ratios.

    SciTech Connect (OSTI)

    Hoffman, E. A.; Yang, W. S.; Hill, R. N.; Nuclear Engineering Division

    2008-05-05T23:59:59.000Z

    A consistent set of designs for 1000 MWt commercial-scale sodium-cooled Advance Burner Reactors (ABR) have been developed for both metal and oxide-fueled cores with conversion ratios from breakeven (CR=1.0) to fertile-free (CR=0.0). These designs are expected to satisfy thermal and irradiation damage limits based on the currently available data. The very low conversion ratio designs require fuel that is beyond the current fuel database, which is anticipated to be qualified by and for the Advanced Burned Test Reactor. Safety and kinetic parameters were calculated, but a safety analysis was not performed. Development of these designs was required to achieve the primary goal of this study, which was to generate representative fuel cycle mass flows for system studies of ABRs as part of the Global Nuclear Energy Partnership (GNEP). There are slight variations with conversion ratio but the basic ABR configuration consists of 144 fuel assemblies and between 9 and 22 primary control assemblies for both the metal and oxide-fueled cores. Preliminary design studies indicated that it is feasible to design the ABR to accommodate a wide range of conversion ratio by employing different assembly designs and including sufficient control assemblies to accommodate the large reactivity swing at low conversion ratios. The assemblies are designed to fit within the same geometry, but the size and number of fuel pins within each assembly are significantly different in order to achieve the target conversion ratio while still satisfying thermal limits. Current irradiation experience would allow for a conversion ratio of somewhat below 0.75. The fuel qualification for the first ABR should expand this experience to allow for much lower conversion ratios and higher bunrups. The current designs were based on assumptions about the performance of high and very high enrichment fuel, which results in significant uncertainty about the details of the designs. However, the basic fuel cycle performance trends such as conversion ratio and mass flow parameters are less sensitive to these parameters and the current results should provide a good basis for static and dynamic system analysis. The conversion ratio is fundamentally a ratio of the macroscopic cross section of U-238 capture to that of TRU fission. Since the microscopic cross sections only change moderately with fuel design and isotopic concentration for the fast reactor, a specific conversion ratio requires a specific enrichment. The approximate average charge enrichment (TRU/HM) is 14%, 21%, 33%, 56%, and 100% for conversion ratios of 1.0, 0.75, 0.50, 0.25, and 0.0 for the metal-fueled cores. The approximate average charge enrichment is 17%, 25%, 38%, 60%, and 100% for conversion ratios of 1.0, 0.75, 0.50, 0.25, and 0.0 for the oxide-fueled core. For the split batch cores, the maximum enrichment will be somewhat higher. For both the metal and oxide-fueled cores, the reactivity feedback coefficients and kinetics parameters seem reasonable. The maximum single control assembly reactivity faults may be too large for the low conversion ratio designs. The average reactivity of the primary control assemblies was increased, which may cause the maximum reactivity of the central control assembly to be excessive. The values of the reactivity coefficients and kinetics parameters show that some values appear to improve significantly at lower conversion ratios while others appear far less favorable. Detailed safety analysis is required to determine if these designs have adequate safety margins or if appropriate design modifications are required. Detailed system analysis data has been generated for both metal and oxide-fueled core designs over the entire range of potential burner reactors. Additional data has been calculated for a few alternative fuel cycles. The systems data has been summarized in this report and the detailed data will be provided to the systems analysis team so that static and dynamic system analyses can be performed.

  15. Conceptual Design of a Reactor Pressure Vessel and its Internals for a HPLWR

    SciTech Connect (OSTI)

    Fischer, Kai [EnBW Kraftwerke AG, Kernkraftwerk Philippsburg, Rheinschanzinsel D-76661 Philippsburg (Germany); Starflinger, Joerg; Schulenberg, Thomas [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2006-07-01T23:59:59.000Z

    A design for the Reactor Pressure Vessel (RPV) and its internals for a HPLWR (High Performance Light Water Reactor) is presented. The RPV has been dimensioned using the pressure vessel code for nuclear power plants in Germany. In order to use conventional vessel materials such as 20 MnMoNi 5 5 (United States: SA 508), the vessel inner wall has to be kept only in contact with coolant at inlet temperature. Therefore, the hot coolant pipe connection from the steam plenum to the outlet is separated from the RPV inner wall using a thermal sleeve. The core inside the vessel rests on a support plate which is connected to the core barrel. The steam plenum is fixed on top of the core using support brackets which are attached to the adjustable steam outlet pipes. This way, the steam plenum rests on the outlet flanges of the lower vessel, while the core barrel is suspended at the closure head flange of the vessel to control thermal expansions between the internals and the RPV and to minimize thermal stresses. Both, inlet and outlet mass flows are separated via C-ring seals to prevent mixing. The control rod guides in the upper plenum are also suspended at the vessel flange and aligned inside the core barrel using centering pins. (authors)

  16. Core design study of a supercritical light water reactor with double row fuel rods

    SciTech Connect (OSTI)

    Zhao, C.; Wu, H.; Cao, L.; Zheng, Y. [School of Nuclear Science and Technology, Xi'an Jiaotong Univ., No. 28, Xianning West Road, Xi'an, ShannXi, 710049 (China); Yang, J.; Zhang, Y. [China Nuclear Power Technology Research Inst., Yitian Road, ShenZhen, GuangDong, 518026 (China)

    2012-07-01T23:59:59.000Z

    An equilibrium core for supercritical light water reactor has been designed. A novel type of fuel assembly with dual rows of fuel rods between water rods is chosen and optimized to get more uniform assembly power distributions. Stainless steel is used for fuel rod cladding and structural material. Honeycomb structure filled with thermal isolation is introduced to reduce the usage of stainless steel and to keep moderator temperature below the pseudo critical temperature. Water flow scheme with ascending coolant flow in inner regions is carried out to achieve high outlet temperature. In order to enhance coolant outlet temperature, the radial power distributions needs to be as flat as possible through operation cycle. Fuel loading pattern and control rod pattern are optimized to flatten power distribution at inner regions. Axial fuel enrichment is divided into three parts to control axial power peak, which affects maximum cladding surface temperature. (authors)

  17. Review of the International Thermonuclear Experimental Reactor (ITER) detailed design report

    SciTech Connect (OSTI)

    NONE

    1997-04-18T23:59:59.000Z

    Dr. Martha Krebs, Director, Office of Energy Research at the US Department of Energy (DOE), wrote to the Fusion Energy Sciences Advisory Committee (FESAC), in letters dated September 23 and November 6, 1996, requesting that FESAC review the International Thermonuclear Experimental Reactor (ITER) Detailed Design Report (DDR) and provide its view of the adequacy of the DDR as part of the basis for the United States decision to enter negotiations with the other interested Parties regarding the terms and conditions for an agreement for the construction, operations, exploitation and decommissioning of ITER. The letter from Dr. Krebs, referred to as the Charge Letter, provided context for the review and a set of questions of specific interest.

  18. Summary of Apollo; A D- sup 3 He tokamak reactor design

    SciTech Connect (OSTI)

    Kulcinski, G.L.; Blanchard, T.P.; El-Guebaly, L.A.; Emert, G.A.; Khater, H.Y.; Maynard, C.W.; Mogahed, E.A.; Santarius, J.F.; Sawan, M.E.; Sviatoslavsky, I.N.; Wittenberg, L.J. (Wisconsin Univ., Madison, WI (United States). Fusion Technology Inst.)

    1992-07-01T23:59:59.000Z

    In this paper, the key features of Apollo, a conceptual D-{sup 3}He tokamak reactor for commercial electricity production, are summarized. The 1000-MW (electric) design utilizes direct conversion of transport, neutron, and bremsstrahlung radiation power. The direct conversion method uses reactants, and the thermal conversion cycle uses an organic coolant. Apollo operates in the first-stability regime, with a major radius of 7.89 m, a peak magnetic field on the toroidal field coils of 19.3 T, a 53-MA plasma current, and a 6.7% beta value. The low neutron production of the D-{sup 3}He fuel cycle greatly reduces the radiation damage rate and allows a full-lifetime first wall and structure made of standard steels with only slight modifications to reduce activation levels.

  19. Stress Analysis of Coated Particle Fuel in the Deep-Burn Pebble Bed Reactor Design

    SciTech Connect (OSTI)

    B. Boer; A. M. Ougouag

    2010-05-01T23:59:59.000Z

    High fuel temperatures and resulting fuel particle coating stresses can be expected in a Pu and minor actinide fueled Pebble Bed Modular Reactor (400 MWth) design as compared to the ’standard’ UO2 fueled core. The high discharge burnup aimed for in this Deep-Burn design results in increased power and temperature peaking in the pebble bed near the inner and outer reflector. Furthermore, the pebble power in a multi-pass in-core pebble recycling scheme is relatively high for pebbles that make their first core pass. This might result in an increase of the mechanical failure of the coatings, which serve as the containment of radioactive fission products in the PBMR design. To investigate the integrity of the particle fuel coatings as a function of the irradiation time (i.e. burnup), core position and during a Loss Of Forced Cooling (LOFC) incident the PArticle STress Analysis code (PASTA) has been coupled to the PEBBED code for neutronics, thermal-hydraulics and depletion analysis of the core. Two deep burn fuel types (Pu with or without initial MA fuel content) have been investigated with the new code system for normal and transient conditions including the effect of the statistical variation of thickness of the coating layers.

  20. Design and optimization of a high thermal flux research reactor via Kriging-based algorithm

    E-Print Network [OSTI]

    Kempf, Stephanie Anne

    2011-01-01T23:59:59.000Z

    In response to increasing demands for the services of research reactors, a 5 MW LEU-fueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic ...

  1. Fusion reactor theory and conceptual design. (Latest citations from the INSPEC: Information Services for the Physics and Engineering Communities database). Published Search

    SciTech Connect (OSTI)

    Not Available

    1992-11-01T23:59:59.000Z

    The bibliography contains citations concerning theoretical and conceptual aspects of fusion reactor physics and designs. A variety of fusion reactors is discussed, including Tokamak, experimental, commercial, tandem mirror, and superconducting magnetic. Topics also include fusion reactor materials, Tokamak devices, blanket design, divertors, fusion plasma production, superconducting magnets, and cryogenic systems. (Contains a minimum of 159 citations and includes a subject term index and title list.)

  2. Analysis and design of a saturable reactor assisted soft-switching full-bridge dc-dc converter

    SciTech Connect (OSTI)

    Hamada, Satoshi (Sansha Electric Manufacturing Co., Ltd., Osaka (Japan)); Nakaoka, Mutsuo (Kobe Univ. (Japan). Dept. of Electrical and Electronics Engineering)

    1994-05-01T23:59:59.000Z

    Analysis and design considerations for a saturable reactor assisted soft-switching full-bridge dc-dc converter are presented. The converter has advantages such as low switching losses with no substantial increase in conduction losses, wide load range, and constant frequency operation. To show how to utilize the analysis results, a 350-W, 500-kHz converter is chosen as a design example. The results are verified experimentally on a prototype converter.

  3. Neutronic Analysis of an Advanced Fuel Design Concept for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Xoubi, Ned [ORNL; Primm, Trent [ORNL; Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK)

    2009-01-01T23:59:59.000Z

    This study presents the neutronic analysis of an advanced fuel design concept for the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) that could significantly extend the current fuel cycle length under the existing design and safety criteria. A key advantage of the fuel design herein proposed is that it would not require structural changes to the present HFIR core, in other words, maintaining the same rated power and fuel geometry (i.e., fuel plate thickness and coolant channel dimensions). Of particular practical importance, as well, is the fact that the proposed change could be justified within the bounds of the existing nuclear safety basis. The simulations herein reported employed transport theory-based and exposure-dependent eigenvalue characterization to help improve the prediction of key fuel cycle parameters. These parameters were estimated by coupling a benchmarked three-dimensional MCNP5 model of the HFIR core to the depletion code ORIGEN via the MONTEBURNS interface. The design of an advanced HFIR core with an improved fuel loading is an idea that evolved from early studies by R. D. Cheverton, formerly of ORNL. This study contrasts a modified and increased core loading of 12 kg of 235U against the current core loading of 9.4 kg. The simulations performed predict a cycle length of 39 days for the proposed fuel design, which represents a 50% increase in the cycle length in response to a 25% increase in fissile loading, with an average fuel burnup increase of {approx}23%. The results suggest that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core's life. Also, the new power distribution is comparable or even improved relative to the current power distribution, displaying lower peak to average fission rate densities across the inner fuel element's centerline and bottom cells. In fact, the fission rate density in the outer fuel element also decreased at these key locations for the proposed design. Overall, it is estimated that the advanced core design could increase the availability of the HFIR facility by {approx}50% and generate {approx}33% more neutrons annually, which is expected to yield sizeable savings during the remaining life of HFIR, currently expected to operate through 2014. This study emphasizes the neutronics evaluation of a new fuel design. Although a number of other performance parameters of the proposed design check favorably against the current design, and most of the core design features remain identical to the reference, it is acknowledged that additional evaluations would be required to fully justify the thermal-hydraulic and thermal-mechanical performance of a new fuel design, including checks for cladding corrosion performance as well as for industrial and economic feasibility.

  4. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2009-05-01T23:59:59.000Z

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep 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 graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The final design phase for the first experiment was completed in September 2008, and the fabrication and assembly of the experiment test train as well as installation and testing of the control and support systems that will monitor and control the experiment during irradiation are being completed in early calendar 2009. The first experiment is scheduled to be ready for insertion in the ATR by April 30, 2009. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and data collection systems.

  5. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

    SciTech Connect (OSTI)

    Michael A. Pope

    2011-10-01T23:59:59.000Z

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

  6. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect (OSTI)

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06T23:59:59.000Z

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are similar to those obtained with the traditional UO2 nuclear fuel.

  7. Nuclear Regulatory Commission Handling of Beyond Design Basis Events for Nuclear Power Reactors

    Broader source: Energy.gov [DOE]

    Presenter: Bill Reckley, Chief, Policy and Support Branch, Japan Lessons-Learned Project Directorate, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission US Nuclear Regulatory Commission

  8. Home Energy Professional Certifications (Fact Sheet)

    SciTech Connect (OSTI)

    Not Available

    2012-11-01T23:59:59.000Z

    As the weatherization and home energy upgrade industries expand and gain recognition, the need for a qualified workforce becomes more apparent. The certification component of the Guidelines project was designed to create meaningful and lasting careers for weatherization workers. Intended for experienced home energy professionals, the four new certifications focus on the most common jobs in the industry: energy auditor, retrofit installer technician, crew leader, and quality control inspector.

  9. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    SciTech Connect (OSTI)

    Trianti, Nuri, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Su'ud, Zaki, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Arif, Idam, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) (Indonesia); Riyana, EkaSapta [Nuclear Energy Regulatory Agency (BAPETEN) (Indonesia)

    2014-09-30T23:59:59.000Z

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  10. Preliminary Design For Conventional and Compact Secondary Heat Exchanger in a Molten Salt Reactor

    SciTech Connect (OSTI)

    Piyush Sabharwall; Mike Patterson; Ali Siahpush; Eung Soo Kim

    2012-07-01T23:59:59.000Z

    The strategic goal of the Advance Reactors such as AHTR is to broaden the environmental and economic benefits of nuclear energy in the United States by producing power to meet growing energy demands and demonstrating its applicability to market sectors not being served by light water reactors

  11. 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

    SciTech Connect (OSTI)

    David W. Nigg; Sean R. Morrell

    2012-09-01T23:59:59.000Z

    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 several obsolete components of the current analytical tool set used for ATR neutronics support. 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). It will also greatly facilitate the LEU conversion effort, since the upgraded computational capabilities are now at a stage where they can be, and in fact have been, used for the required physics analysis from the beginning. In this context, extensive scoping neutronics analyses were completed for six preconceptual candidate LEU fuel element designs for the ATR (and for its companion critical facility, ATRC). Of these, four exhibited neutronics performance in what is believed to be an acceptable range. However, there are currently some concerns with regard to fabricability and mechanical performance that have emerged for one of the four latter concepts. Thus three concepts have been selected for more comprehensive conceptual design analysis during the upcoming fiscal year.

  12. Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source

    SciTech Connect (OSTI)

    Gehin, J.C.; Worley, B.A.; Renier, J.P. [Oak Ridge National Lab., TN (United States); Wemple, C.A.; Jahshan, S.N.; Ryskammp, J.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-08-01T23:59:59.000Z

    This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific time steps over the fuel cycle. The Monte Carlo analysis was carried out using the MCNP continuous-energy code, and the few- group diffusion theory calculations were performed using the VENTURE and PDQ code systems. The MCNP code was used primarily for its capability to model the reflector components in realistic geometries as well as the inherent circumvention of cross-section processing requirements and use of energy-collapsed cross sections. The MCNP code was used for evaluations of reflector component reactivity effects and of heat loads in these components. The code was also used as a benchmark comparison against the diffusion-theory estimates of key reactor parameters such as region fluxes, control rod worths, reactivity coefficients, and material worths. The VENTURE and PDQ codes were used to provide independent evaluations of burnup effects, power distributions, and small perturbation worths. The performance and safety calculations performed over the subject time period are summarized, and key results are provided. The key results include flux and power distributions over the fuel cycle, silicon production rates, fuel burnup rates, component reactivities, control rod worths, component heat loads, shutdown reactivity margins, reactivity coefficients, and isotope production rates.

  13. Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 39, No. 11, p. 11691181 (November 2002) Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M"

    E-Print Network [OSTI]

    Laughlin, Robert B.

    Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 39, No. 11, p. 1169­1181 (November 2002) Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M" Mitsuru KAMBE Central Research Institute and accepted September 10, 2002) A metal fueled modular island core sodium cooled fast breeder reactor concept

  14. Status of EC solid breeder blanket designs and R&D for DEMO fusion reactors

    SciTech Connect (OSTI)

    Dalle Donne, M. [INR, Karlsruhe (Russian Federation); Anziedi, L.A. [C.R.E., Franscati (Italy); Kwast, H. [ECN, Petten (Netherlands)] [and others

    1994-12-31T23:59:59.000Z

    In the framework of the European Community Fusion Technology Program four blanket concepts for a DEMO reactor are being investigated. DEMO is the next step after ITER. It should ensure tritium self-sufficiency and operate at coolant temperatures high enough to have a reasonable plant efficiency. Further requirements have been specified for the four concepts, namely an average neutron wall load of 2.2 MW/m{sup 2}, a blanket lifetime of 20000 hours and the capability of the blanket segment to withstand the forces caused by a rapid distribution of the plasma current (20 MA to zero in 20 ms), so that after the disruption the segment can still allow a comparison of the various options, in view of reducing this number to two in 1995 and to design and develop modules and articles representative of the chosen blankets to be tested in ITER. The present paper deals with two solid breeder concepts. They have many features in common: both use high pressure helium as coolant and helium to purge the tritium from the breeder material, martensitic steel as structural material and beryllium as neutron multiplier. The configuration of the two blankets are however different: in the B.I.T. (Breeder Inside Tube) concept the breeder material is LiAlO{sub 2} or LiZrO{sub 3} in the form of annular pellets contained in tubes surrounded by beryllium blocks, the coolant helium being outside the tubes, whereas in the B.O.T. (Breeder out of Tube) the breeder and multiplier material are Li{sub 4}SiO{sub 4} and beryllium pebbles forming a mixed bed placed outside the tubes containing the coolant helium.

  15. Stability analysis of the boiling water reactor : methods and advanced designs

    E-Print Network [OSTI]

    Hu, Rui, Ph. D. Massachusetts Institute of Technology

    2010-01-01T23:59:59.000Z

    Density Wave Oscillations (DWOs) are known to be possible when a coolant undergoes considerable density reduction while passing through a heated channel. In the development of boiling water reactors (BWRs), there has been ...

  16. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    type fast reactor of the IV generation for regional powerELECTRA-FCC: a centre for Generation IV system research andunder the framework of generation-IV nuclear pro- grams or

  17. Design, analysis and optimization of the power conversion system for the Modular Pebble Bed Reactor System

    E-Print Network [OSTI]

    Wang, Chunyun, 1968-

    2003-01-01T23:59:59.000Z

    The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a GenIV nuclear system. The availability of controllable ...

  18. Design strategies for optimizing high burnup fuel in pressurized water reactors

    E-Print Network [OSTI]

    Xu, Zhiwen, 1975-

    2003-01-01T23:59:59.000Z

    This work is focused on the strategy for utilizing high-burnup fuel in pressurized water reactors (PWR) with special emphasis on the full array of neutronic considerations. The historical increase in batch-averaged discharge ...

  19. An inverted pressurized water reactor design with twisted-tape swirl promoters

    E-Print Network [OSTI]

    Nguyen, Nghia T. (Nghia Tat)

    2014-01-01T23:59:59.000Z

    An Inverted Fuel Pressurized Water Reactor (IPWR) concept was previously investigated and developed by Paolo Ferroni at MIT with the effort to improve the power density and capacity of current PWRs by modifying the core ...

  20. Optimized core design of a supercritical carbon dioxide-cooled fast reactor

    E-Print Network [OSTI]

    Handwerk, Christopher S. (Christopher Stanley), 1974-

    2007-01-01T23:59:59.000Z

    Spurred by the renewed interest in nuclear power, Gas-cooled Fast Reactors (GFRs) have received increasing attention in the past decade. Motivated by the goals of the Generation-IV International Forum (GIF), a GFR cooled ...

  1. Thermal hydraulic design and analysis of a large lead-cooled reactor with flexible conversion ratio

    E-Print Network [OSTI]

    Nikiforova, Anna S., S.M. Massachusetts Institute of Technology

    2008-01-01T23:59:59.000Z

    This thesis contributes to the Flexible Conversion Ratio Fast Reactor Systems Evaluation Project, a part of the Nuclear Cycle Technology and Policy Program funded by the Department of Energy through the Nuclear Energy ...

  2. Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs

    SciTech Connect (OSTI)

    Yoder, G.L.

    2005-10-03T23:59:59.000Z

    This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.

  3. Building Operator Certification

    E-Print Network [OSTI]

    Lilley, D.

    2013-01-01T23:59:59.000Z

    Building Operator Certification Energy Efficiency through Operator Training CATEE December 18, 2013 – San Antonio, TX Dennis Lilley, CEM, PMP ESL-KT-13-12-49 CATEE 2013: Clean Air Through Energy Efficiency Conference, San Antonio, Texas Dec. 16...-18 Building Operator Certification Energy Efficiency through Operator Training What is Building Operator Certification? Industry-recognized credential in energy efficient building operation practices Created with 100 industry experts Launched in 1996 9...

  4. Certification Status Search Tool

    E-Print Network [OSTI]

    Certification Status Search Tool (SB/DVBE) Illustrated User Manual #12;2 Certification Status Search Tool (SB/DVBE) Rev. 5/14/09 Copyright © 2008 RFP Depot, LLC dba BidSync All Rights Reserved://www.bidsync.com #12;3 Certification Status Search Tool (SB/DVBE) Rev. 5/14/09 Introduction BidSync is a powerful e

  5. Pressure loadings of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) reactor release mitigation structures from large-break LOCAs

    SciTech Connect (OSTI)

    Sienicki, J.J.; Horak, W.C. (Argonne National Lab., IL (USA); Brookhaven National Lab., Upton, NY (USA))

    1989-01-01T23:59:59.000Z

    Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs.

  6. National Certification Standard

    Broader source: Energy.gov (indexed) [DOE]

    Certification Standard for the Geothermal Heat Pump Industry Principal Investigator John Kelly Geothermal Heat Pump Consortium GSHP Demonstration Projects May 18, 2010 This...

  7. An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

    SciTech Connect (OSTI)

    Farzad Rahnema; Dingkang Zhang; Abderrafi Ougouag; Frederick Gleicher

    2011-04-04T23:59:59.000Z

    The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulators

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

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01T23:59:59.000Z

    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 significantly different from the first two experiments, though the control and monitoring systems are very similar. The purpose and design of this experiment will be discussed followed by its progress and status to date.

  9. Th/U-233 multi-recycle in pressurized water reactors : feasibility study of multiple homogeneous and heterogeneous assembly designs.

    SciTech Connect (OSTI)

    Yun, D.; Taiwo, T. A.; Kim, T. K.; Mohamed, A.; Nuclear Engineering Division

    2010-10-01T23:59:59.000Z

    The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle. The possibility for thorium utilization in a multi-recycle system has also been considered in past literature, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this study is to evaluate the potential of Th/U-233 fuel multi-recycle in current LWRs, focusing on pressurized water reactors (PWRs). Approaches for sustainable multi-recycle without the need for external fissile material makeup have been investigated. The intent is to obtain a design that allows existing PWRs to be used with minimal modifications.

  10. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    SciTech Connect (OSTI)

    Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

    2011-02-01T23:59:59.000Z

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  11. Browser Interfaces and EV-SSL Certificates: Confusion, Inconsistencies and HCI Challenges

    E-Print Network [OSTI]

    Van Oorschot, Paul

    Browser Interfaces and EV-SSL Certificates: Confusion, Inconsistencies and HCI Challenges Jennifer (EV) SSL certificates has caused web browser manufacturers to take a new look at how they design SSL certificates rather than in- creasing trust. We perform a systematic walkthrough involving

  12. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    SciTech Connect (OSTI)

    Not Available

    1980-09-01T23:59:59.000Z

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases.

  13. Flow Components in a NaK Test Loop Designed to Simulate Conditions in a Nuclear Surface Power Reactor

    SciTech Connect (OSTI)

    Polzin, Kurt A.; Godfroy, Thomas J. [NASA Marshall Space Flight Center Propulsion Research and Technology Applications Branch/ER24, MSFC, AL 35812 (United States)

    2008-01-21T23:59:59.000Z

    A test loop using NaK as the working fluid is presently in use to study material compatibility effects on various components that comprise a possible nuclear reactor design for use on the lunar surface. A DC electromagnetic (EM) pump has been designed and implemented as a means of actively controlling the NaK flow rate through the system and an EM flow sensor is employed to monitor the developed flow rate. These components allow for the matching of the flow rate conditions in test loops with those that would be found in a full-scale surface-power reactor. The design and operating characteristics of the EM pump and flow sensor are presented. In the EM pump, current is applied to a set of electrodes to produce a Lorentz body force in the fluid. A measurement of the induced voltage (back-EMF) in the flow sensor provides the means of monitoring flow rate. Both components are compact, employing high magnetic field strength neodymium magnets thermally coupled to a water-cooled housing. A vacuum gap limits the heat transferred from the high temperature NaK tube to the magnets and a magnetically-permeable material completes the magnetic circuit. The pump is designed to produce a pressure rise of 34.5 kPa, and the flow sensor's predicted output is roughly 20 mV at the loop's nominal flow rate of 0.114 m{sup 3}/hr.

  14. Exploratory design study on reactor configurations for carbon dioxide capture from conventional power plants employing regenerable solid sorbents

    SciTech Connect (OSTI)

    Yang, W.C.; Hoffman, J. [US DOE, Pittsburgh, PA (USA). National Energy Technology Laboratory

    2009-01-15T23:59:59.000Z

    Preliminary commercial designs were carried out for a fluidized bed as a CO{sub 2} adsorber and a moving bed as a CO{sub 2} regenerator. Reverse engineering methodology was employed on the basis of a commercial 500 MW supercritical PC power plant whereby the boundaries required for a particular reactor design and configuration could be set. Employing the proposed moving bed for regenerator is, however, not promising because of poor heat transfer, evolution of CO{sub 2} during regeneration, and high pressure drop when small particles are used. If regeneration kinetics is as slow as reported in tens of minutes, the bed height can be quite high and the reactor can be quite costly. In its place, a so-called assisted self-fluidization bed with embedded heat transfer surface was proposed. Theoretically, there is no reason why the fluidized bed cannot be successfully designed and operated both as an adsorber and a regenerator under proper adsorption and regeneration kinetics. Recent publications, where fluidized beds, circulating fluidized beds, or a combination of them were employed both as an adsorber and a regenerator, were cited. Staging may not be necessary employing the fluidized bed technology because of the capability to control reaction temperature at the optimum operating temperature through embedded heat transfer surface in the fluidized beds. Even if the staging is necessary, the implementation of staging in fluidized beds at ambient pressure and moderate temperature is relatively easy and with minimum cost penalty. Example designs are presented.

  15. Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

    SciTech Connect (OSTI)

    Ronald G. Ballinger Chunyun Wang Andrew Kadak Neil Todreas

    2004-08-30T23:59:59.000Z

    The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R&D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%.

  16. Experimental study of Siphon breaker about size effect in real scale reactor design

    SciTech Connect (OSTI)

    Kang, S. H. [Mechanical Engineering Dept., POSTECH, Pohang, 790-784 (Korea, Republic of); Ahn, H. S. [Div. of Advanced Nuclear Engineering, POSTECH, Pohang, 790-784 (Korea, Republic of); Kim, J. M. [Mechanical Engineering Dept., POSTECH, Pohang, 790-784 (Korea, Republic of); Joo, H. M. [Dept. of Nuclear Engineering, Hanyang Univ., Seoul, 133-791 (Korea, Republic of); Lee, K. Y.; Seo, K.; Chi, D. Y. [KAERI, Yuseong, Daejeon, 305-353 (Korea, Republic of); Kim, M. H. [Div. of Advanced Nuclear Engineering, POSTECH, Pohang, 790-784 (Korea, Republic of)

    2012-07-01T23:59:59.000Z

    Rupture accident within the pipe of a nuclear reactor is one of the main causes of a loss of coolant accident (LOCA). Siphon-breaking is a passive method that can prevent a LOCA. In this study, either a line or a hole is used as a siphon-breaker, and the effect of various parameters, such as the siphon-breaker size, pipe rupture point, pipe rupture size, and the presence of an orifice, are investigated using an experimental facility similar in size to a full-scale reactor. (authors)

  17. Export Compliance Certificate Program

    E-Print Network [OSTI]

    Barrett, Jeffrey A.

    Export Compliance Certificate Program Accelerate Your Career BusinessandManagement extension.uci.edu/export bearing the UC seal signifies a well- known, uncompromising standard of academic excellence. #12;Export Compliance Certificate Program The importance of understanding export controls and how to develop

  18. Business Analyst Certificate Program

    E-Print Network [OSTI]

    Rose, Michael R.

    Business Analyst Certificate Program BusinessandManagement extension.uci.edu/ba #12;Business Analyst Certificate Program Business Analysts Capture Requirements to Build What the Customer Wants. The Business Analyst serves as the key liaison between the client, stakeholders, and the solutions team

  19. Graduate Certificate in Transportation

    E-Print Network [OSTI]

    Bertini, Robert L.

    Graduate Certificate in Transportation Nohad A. Toulan School of Urban Studies and Planning of Engineering and Computer Science integrated transportation systems. The Graduate Certificate in Transportation their capabilities. Students in the program can choose among a wide range of relevant courses in transportation

  20. Scandinavian Studies Certificate

    E-Print Network [OSTI]

    Saldin, Dilano

    , green energy, sustainability, and information technology. To complete the certificate, students must, and Sweden. Certificate programs are similar to minors. They are comprised of 15 to 26 credits University, UWM's sister university in Sweden. Students attending Malardalen University pay UWM tuition

  1. ORACLE CERTIFICATION Oracle Database

    E-Print Network [OSTI]

    Loudon, Catherine

    ORACLE CERTIFICATION Oracle Database Administration Certificate Program Train with the best. Get your Oracle Database Administration education from the number-one provider* of Oracle training-on, lab-based understanding of Oracle, the world's leading database platform, and long the product

  2. IBM Certification White Paper

    E-Print Network [OSTI]

    surrounding IT certification vary depending on geography, company strategy and the product it supports by individuals and the companies that employ them as a differentiator in the IT Industry. As competition, certification offers several other advantages for both individuals and companies including: · Increased income

  3. Conceptual design of thorium-fuelled Mitrailleuse accelerator-driven subcritical reactor using D-Be neutron source

    SciTech Connect (OSTI)

    Kokubo, Y. [Quan Japan Company Limited, 3-9-15 Sannomiya-cho, Chuo-ku, Kobe, Hyogo, 650-0021 (Japan); Kamei, T. [Research Inst. for Applied Sciences, 49 Tanaka Ohicho, Sakyo-ku, Kyoto-shi, Kyoto, 606-8202 (Japan)

    2012-07-01T23:59:59.000Z

    A distributed accelerator is a charged-particle accelerator that uses a new acceleration method based on repeated electrostatic acceleration. This method offers outstanding benefits not possible with the conventional radio-frequency acceleration method, including: (1) high acceleration efficiency, (2) large acceleration current, and (3) lower failure rate made possible by a fully solid-state acceleration field generation circuit. A 'Mitrailleuse Accelerator' is a product we have conceived to optimize this distributed accelerator technology for use with a high-strength neutron source. We have completed the conceptual design of a Mitrailleuse Accelerator and of a thorium-fuelled subcritical reactor driven by a Mitrailleuse Accelerator. This paper presents the conceptual design details and approach to implementing the subcritical reactor core. We will spend the next year or so on detailed design work, and then will start work on developing a prototype for demonstration. If there are no obstacles in setting up a development organization, we expect to finish verifying the prototype's performance by the third quarter of 2015. (authors)

  4. Design and Testing of a 10B4C Capsule for Spectral-Tailoring in Mixed-Spectrum Reactors

    SciTech Connect (OSTI)

    Greenwood, Lawrence R.; Wittman, Richard S.; Metz, Lori A.; Finn, Erin C.; Friese, Judah I.

    2014-04-11T23:59:59.000Z

    A boron carbide capsule highly enriched in 10B has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. New experiments show that enriching the boron to 96% B-10 results in additional absorption of neutrons in the resonance region thereby producing a neutron spectrum that is much closer to a pure 235U fission spectrum. A cadmium outer cover was used to reduce thermal heating. The neutron spectrum calculated with MCNP was found to be in very good agreement with measured activation rates from neutron fluence monitors.

  5. Certification plan for safety and PRA codes

    SciTech Connect (OSTI)

    Toffer, H.; Crowe, R.D. (Westinghouse Hanford Co., Richland, WA (United States)); Ades, M.J. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1990-05-01T23:59:59.000Z

    A certification plan for computer codes used in Safety Analyses and Probabilistic Risk Assessment (PRA) for the operation of the Savannah River Site (SRS) reactors has been prepared. An action matrix, checklists, and a time schedule have been included in the plan. These items identify what is required to achieve certification of the codes. A list of Safety Analysis and Probabilistic Risk Assessment (SA PRA) computer codes covered by the certification plan has been assembled. A description of each of the codes was provided in Reference 4. The action matrix for the configuration control plan identifies code specific requirements that need to be met to achieve the certification plan's objectives. The checklist covers the specific procedures that are required to support the configuration control effort and supplement the software life cycle procedures based on QAP 20-1 (Reference 7). A qualification checklist for users establishes the minimum prerequisites and training for achieving levels of proficiency in using configuration controlled codes for critical parameter calculations.

  6. Certification plan for safety and PRA codes

    SciTech Connect (OSTI)

    Toffer, H.; Crowe, R.D. [Westinghouse Hanford Co., Richland, WA (United States); Ades, M.J. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1990-05-01T23:59:59.000Z

    A certification plan for computer codes used in Safety Analyses and Probabilistic Risk Assessment (PRA) for the operation of the Savannah River Site (SRS) reactors has been prepared. An action matrix, checklists, and a time schedule have been included in the plan. These items identify what is required to achieve certification of the codes. A list of Safety Analysis and Probabilistic Risk Assessment (SA&PRA) computer codes covered by the certification plan has been assembled. A description of each of the codes was provided in Reference 4. The action matrix for the configuration control plan identifies code specific requirements that need to be met to achieve the certification plan`s objectives. The checklist covers the specific procedures that are required to support the configuration control effort and supplement the software life cycle procedures based on QAP 20-1 (Reference 7). A qualification checklist for users establishes the minimum prerequisites and training for achieving levels of proficiency in using configuration controlled codes for critical parameter calculations.

  7. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    DOE Patents [OSTI]

    Crawford, Douglas C. (Idaho Falls, ID); Porter, Douglas L. (Idaho Falls, ID); Hayes, Steven L. (Idaho Falls, ID); Hill, Robert N. (Bolingbrook, IL)

    1999-01-01T23:59:59.000Z

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both.

  8. Application-specific integrated circuit design for a typical pressurized water reactor pressure channel trip

    SciTech Connect (OSTI)

    Battle, R.E.; Manges, W.W.; Emery, M.S.; Vendermolen, R.I. [Oak Ridge National Lab., TN (United States); Bhatt, S. [Electric Power Research Inst., Palo Alto, CA (United States)

    1994-03-01T23:59:59.000Z

    This article discusses the use of application-specific integrated circuits (ASICs) in nuclear plant safety systems. ASICs have certain advantages over software-based systems because they can be simple enough to be thoroughly tested, and they can be tailored to replace existing equipment. An architecture to replace a pressurized water reactor pressure channel trip is presented. Methods of implementing digital algorithms are also discussed.

  9. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    SciTech Connect (OSTI)

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

    1997-12-01T23:59:59.000Z

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr-Hf alloy or an alloy of Pu-Zr-Hf or a combination of both.

  10. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    DOE Patents [OSTI]

    Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

    1999-03-23T23:59:59.000Z

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

  11. Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels -- Final Report under the International Nuclear Energy Research Initiative (I-NERI)

    SciTech Connect (OSTI)

    David Petti; Philippe Martin; Mayeul Phélip; Ronald Ballinger; Petti does not have NT account

    2004-12-01T23:59:59.000Z

    The objective of this INERI project was to develop improved fuel behavior models for gas reactor coated-particle fuels and to explore improved coated-particle fuel designs that could be used reliably at very high burnups and potentially in gas-cooled fast reactors. Project participants included the Idaho National Engineering Laboratory (INEEL), Centre Étude Atomique (CEA), and the Massachusetts Institute of Technology (MIT). To accomplish the project objectives, work was organized into five tasks.

  12. Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant

    SciTech Connect (OSTI)

    Chang H. Oh; Eung Soo Kim; Steven Sherman

    2008-04-01T23:59:59.000Z

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood.

  13. EA-1991: 10 CFR 433 and 10 CFR 435: Green Building Certification...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Standards for the Design and Construction of New Federal Low-Rise Residential Buildings" (Green Building Certification Program) October 14, 2014 EA-1991: Final Environmental...

  14. Collective, Collaborative or Competitive? An Analysis of EPR Approaches using Material Recovery Certificates for

    E-Print Network [OSTI]

    Gutowski, Timothy

    Massachusetts Institute of Technology, Laboratory for Manufacturing and Productivity, 77, Massachusetts Avenue, WEEE Recycling, Material Recovery Certificates, Regulation Design 1. The European EPR system landscape

  15. Certification process of safety analysis and risk management computer codes at the Savannah River Site

    SciTech Connect (OSTI)

    Ades, M.J. [Westinghouse Savannah River Co., Aiken, SC (United States); Toffer, H.; Lewis, C.J.; Crowe, R.D. [Westinghouse Hanford Co., Richland, WA (United States)

    1992-05-01T23:59:59.000Z

    The commitment by Westinghouse Savannah River Company (WSRC) to bring safety analysis and risk management codes into compliance with national and sitewide quality assurance requirements necessitated a systematic, structured approach. As a part of this effort, WSRC, in cooperation with the Westinghouse Hanford Company, has developed and implemented a certification process for the development and control of computer software. Safety analysis and risk management computer codes pertinent to reactor analyses were selected for inclusion in the certification process. As a first step, documented plans were developed for implementing verification and validation of the codes, and establishing configuration control. User qualification guidelines were determined. The plans were followed with an extensive assessment of the codes with respect to certification status. Detailed schedules and work plans were thus determined for completing certification of the codes considered. Although the software certification process discussed is specific to the application described, it is sufficiently general to provide useful insights and guidance for certification of other software.

  16. Certification process of safety analysis and risk management computer codes at the Savannah River Site

    SciTech Connect (OSTI)

    Ades, M.J. (Westinghouse Savannah River Co., Aiken, SC (United States)); Toffer, H.; Lewis, C.J.; Crowe, R.D. (Westinghouse Hanford Co., Richland, WA (United States))

    1992-01-01T23:59:59.000Z

    The commitment by Westinghouse Savannah River Company (WSRC) to bring safety analysis and risk management codes into compliance with national and sitewide quality assurance requirements necessitated a systematic, structured approach. As a part of this effort, WSRC, in cooperation with the Westinghouse Hanford Company, has developed and implemented a certification process for the development and control of computer software. Safety analysis and risk management computer codes pertinent to reactor analyses were selected for inclusion in the certification process. As a first step, documented plans were developed for implementing verification and validation of the codes, and establishing configuration control. User qualification guidelines were determined. The plans were followed with an extensive assessment of the codes with respect to certification status. Detailed schedules and work plans were thus determined for completing certification of the codes considered. Although the software certification process discussed is specific to the application described, it is sufficiently general to provide useful insights and guidance for certification of other software.

  17. Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design

    SciTech Connect (OSTI)

    Smith, C

    2010-02-22T23:59:59.000Z

    The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

  18. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

    uranium (LEU) cores. Unlike light water reactors (LWRs), the ultimate heat sink for decay heat removal

  19. Design of central irradiation facilities for the MITR-II research reactor

    E-Print Network [OSTI]

    Meagher, Paul Christopher

    1976-01-01T23:59:59.000Z

    Design analysis studies have been made for various in-core irradiation facility designs which are presently used, or proposed for future use in the MITR-II. The information obtained includes reactivity effects, core flux ...

  20. 401 Certification (Vermont)

    Broader source: Energy.gov [DOE]

    A 401 Certification is required for any project that triggers a federal permit or license. Examples include licenses from the Nuclear Regulatory Commission and any project which requires a permit...

  1. SHIELD certification package

    SciTech Connect (OSTI)

    Boman, C.

    1992-02-01T23:59:59.000Z

    Certification as applied to existing computer codes includes the verification and validation process, placing the code in configuration control, establishing user qualification standards and training requirements. All software intended for use in critical calculations must be certified. This report is intended to fulfill the requirements for the certification of the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system built February, 1992, by W.S. Parks. These modules are used for burnup, cooling, separate, and edit calculations.

  2. Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor

    SciTech Connect (OSTI)

    Piyush Sabharwall; Ali Siahpush; Michael McKellar; Michael Patterson; Eung Soo Kim

    2012-06-01T23:59:59.000Z

    The goals of next generation nuclear reactors, such as the high temperature gas-cooled reactor and advance high temperature reactor (AHTR), are to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. The need for efficiency, compactness, and safety challenge the boundaries of existing heat exchanger technology, giving rise to the following study. Various studies have been performed in attempts to update the secondary heat exchanger that is downstream of the primary heat exchanger, mostly because its performance is strongly tied to the ability to employ more efficient conversion cycles, such as the Rankine super critical and subcritical cycles. This study considers two different types of heat exchangers—helical coiled heat exchanger and printed circuit heat exchanger—as possible options for the AHTR secondary heat exchangers with the following three different options: (1) A single heat exchanger transfers all the heat (3,400 MW(t)) from the intermediate heat transfer loop to the power conversion system or process plants; (2) Two heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants, each exchanger transfers 1,700 MW(t) with a parallel configuration; and (3) Three heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants. Each heat exchanger transfers 1,130 MW(t) with a parallel configuration. A preliminary cost comparison will be provided for all different cases along with challenges and recommendations.

  3. Reactor design for uniform chemical vapor deposition-grown films without substrate rotation

    SciTech Connect (OSTI)

    Wanlass, Mark (Golden, CO)

    1987-01-01T23:59:59.000Z

    A quartz reactor vessel for growth of uniform semiconductor films includes a vertical, cylindrical reaction chamber in which a substrate-supporting pedestal provides a horizontal substrate-supporting surface spaced on its perimeter from the chamber wall. A cylindrical confinement chamber of smaller diameter is disposed coaxially above the reaction chamber and receives reaction gas injected at a tangent to the inside chamber wall, forming a helical gas stream that descends into the reaction chamber. In the reaction chamber, the edge of the substrate-supporting pedestal is a separation point for the helical flow, diverting part of the flow over the horizontal surface of the substrate in an inwardly spiraling vortex.

  4. Reactor design for uniform chemical vapor deposition-grown films without substrate rotation

    DOE Patents [OSTI]

    Wanlass, M.

    1985-02-19T23:59:59.000Z

    A quartz reactor vessel for growth of uniform semiconductor films includes a vertical, cylindrical reaction chamber in which a substrate-supporting pedestal provides a horizontal substrate-supporting surface spaced on its perimeter from the chamber wall. A cylindrical confinement chamber of smaller diameter is disposed coaxially above the reaction chamber and receives reaction gas injected at a tangent to the inside chamber wall, forming a helical gas stream that descends into the reaction chamber. In the reaction chamber, the edge of the substrate-supporting pedestal is a separation point for the helical flow, diverting part of the flow over the horizontal surface of the substrate in an inwardly spiraling vortex.

  5. certification, compliance and enforcement regulations for Commercial...

    Office of Environmental Management (EM)

    certification, compliance and enforcement regulations for Commercial Refrigeration Equipment (CRE) certification, compliance and enforcement regulations for Commercial...

  6. Clinch River breeder reactor sodium fire protection system design and development

    SciTech Connect (OSTI)

    Foster, K.W.; Boasso, C.J.; Kaushal, N.N.

    1984-04-13T23:59:59.000Z

    To assure the protection of the public and plant equipment, improbable accidents were hypothesized to form the basis for the design of safety systems. One such accident is the postulated failure of the Intermediate Heat Transfer System (IHTS) piping within the Steam Generator Building (SGB), resulting in a large-scale sodium fire. This paper discusses the design and development of plant features to reduce the consequences of the accident to acceptable levels. Additional design solutions were made to mitigate the sodium spray contribution to the accident scenario. Sodium spill tests demonstrated that large sodium leaks can be safely controlled in a sodium-cooled nuclear power plant.

  7. acid immunoaffinity reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    sludge blanket reactors (more) Ning, Zuojun. 2009-01-01 8 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  8. Preventing fuel failure for a beyond design basis accident in a fluoride salt cooled high temperature reactor

    E-Print Network [OSTI]

    Minck, Matthew J. (Matthew Joseph)

    2013-01-01T23:59:59.000Z

    The fluoride salt-cooled high-temperature reactor (FHR) combines high-temperature coated-particle fuel with a high-temperature salt coolant for a reactor with unique market and safety characteristics. This combination can ...

  9. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    SciTech Connect (OSTI)

    Monado, Fiber, E-mail: fiber.monado@gmail.com [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Aziz, Ferhat [National Nuclear Energy Agency of Indonesia (BATAN) (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-02-12T23:59:59.000Z

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  10. Design of compact intermediate heat exchangers for gas cooled fast reactors

    E-Print Network [OSTI]

    Gezelius, Knut, 1978-

    2004-01-01T23:59:59.000Z

    Two aspects of an intermediate heat exchanger (IHX) for GFR service have been investigated: (1) the intrinsic characteristics of the proposed compact printed circuit heat exchanger (PCHE); and (2) a specific design optimizing ...

  11. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    SciTech Connect (OSTI)

    Riley, D.R.; Dickson, P.W.

    1981-01-01T23:59:59.000Z

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle.

  12. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

    Reactor. in Proceedings of ICAPP’10 (American NuclearFHR. in Proceedings of ICAPP’12 (American Nuclear Society,

  13. Designing an enzymatic oscillator: Bistability and feedback controlled oscillations with glucose oxidase in a continuous flow stirred tank reactor

    E-Print Network [OSTI]

    Epstein, Irving R.

    oxidase in a continuous flow stirred tank reactor Vladimir K. Vanag,a David G. Míguez,b and Irving R as the flow rate is varied in a continuous flow stirred tank reactor. Oscillations in pH can be obtained experiments, this feedback consists of an inflow of hydroxide ion at a rate that depends on H+ in the reactor

  14. RAMI Analysis for Designing and Optimizing Tokamak Cooling Water System (TCWS) for the ITER's Fusion Reactor

    SciTech Connect (OSTI)

    Ferrada, Juan J [ORNL] [ORNL; Reiersen, Wayne T [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

    U.S.-ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment. TCWS transfers heat generated in the Tokamak during nominal pulsed operation - 850 MW at up to 150 C and 4.2 MPa water pressure. Impurities are diffused from in-vessel components and the vacuum vessel by water baking at 200-240 C at up to 4.4 MPa. TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER's success and interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation, each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements. RAMI analysis will indicate appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. This analysis helps demonstrate that using proven, commercially available equipment is better than using custom-designed equipment with no field experience and lowers specific costs while providing higher reliability. This paper presents a brief description of the TCWS conceptual design and the application of RAMI tools to optimize the design at different stages during the project.

  15. State-of-the-art review and report on critical aspects and scale-up considerations in the design of fluidized-bed reactors. Final report on Phase 1

    SciTech Connect (OSTI)

    Not Available

    1980-01-01T23:59:59.000Z

    Information is given on the design of distributor plates and opening geometry to provide uniform flow over the reactor area. The design of granular bed filters is also considered. Pressure drops and particle size in the bed are discussed. (LTN)

  16. Reactor design for uniform chemical vapor deposition-grown films without substrate rotation

    SciTech Connect (OSTI)

    Wanlass, M.

    1987-03-17T23:59:59.000Z

    A reactor vessel is described for chemical vapor deposition of a uniform semiconductor film on a substrate, comprising: a generally cylindrical reaction chamber for receiving a substrate and a flow of reaction gas capable of depositing a film on the substrate under the conditions of the chamber, the chamber having upper and lower portion and being oriented about a vertical axis; a supporting means having a substrate support surface generally perpendicular to the vertical axis for carrying the substrate within the lower portion of the reaction chamber in a predetermined relative position with respect to the upper portion of the reaction chamber, the upper portion including a cylindrically shaped confinement chamber. The confinement chamber has a smaller diameter than the lower portion of the reaction chamber and is positioned above the substrate support surface; and a means for introducing a reaction gas into the confinement chamber in a nonaxial direction so as to direct the reaction gas into the lower portion of the reaction chamber with a non-axial flow having a rotational component with respect to the vertical axis. In this way the reaction gas defines an inward vortex flow pattern with respect to the substrate surface.

  17. Packed-bed reactor/silent-discharge plasma design data report

    SciTech Connect (OSTI)

    NONE

    1996-05-01T23:59:59.000Z

    In 1992, Congress passed the Federal Facility Compliance Act requiring the U.S. Department of Energy (DOE) to treat and dispose of its mixed waste in accordance with Resource Conservation and Recovery Act (RCRA) land disposal restrictions (LDRs). The DOE Albuquerque Operations Office (AL) currently does not have adequate systems to treat the mixed wastes generated and stored at the nine DOE-AL sites. In response to the need for mixed-waste treatment capacity, DOE-AL organized a Treatment Selection Team under the Mixed-Waste Treatment Program (MWTP) to match mixed wastes with treatment options and develop a strategy for treatment of its mixed waste. The strategy developed by the Treatment Selection Team, as described in the AL Mixed-Waste Treatment Plan (DOE 1994), is to use available off-site commercial treatment facilities for all wastes that can be successfully and cost-effectively treated by such facilities. Where no appropriate commercial treatment facilities exist, mobile treatment units (MTUs) would be developed to treat wastes at the sites where the wastes are generated. Treatment processes used for mixed waste must not only address the hazardous component (i.e., meet LDRs) but also must contain the radioactive component in a form that allows final disposal while protecting workers, the public, and the environment. The packed-bed reactor/silent discharge plasma was chosen as a potential candidate for the treatment of the mixed wastes. The process is described.

  18. Design considerations for a steady state fusion reactor's thermal energy dump (TED) with emphasis on SAFFIRE

    SciTech Connect (OSTI)

    Werley, K.A.

    1980-01-01T23:59:59.000Z

    This work examines the use of a thermal dump to handle the severe particle and energy handling requirements of a diverted plasma. We outline a general approach for evaluating the design parameters and limitations of a thermal dump, considering such things as thermomechanical and erosion effects, compatibility, availability, machinability, coolant recirculation, vacuum pumping, economics, lifetime, etc. To demonstrate how the performance requirements are reflected in design decisions, we apply a solid-walled dump to a small-sized field reversed mirror (FRM). We also examine a liquid-lithium droplet thermal dump and point out some distinct advantages of this new concept over the solid-wall design in reducing stress, erosion, and vacuum pumping problems. The chief disadvantages of this scheme include liquid-metal safe-handling problems, vapor pressure-temperature limitations, and the need for differential pumping if T/sub Li/ > 310/sup 0/C is desired.

  19. Linear ParameterVarying versus Linear TimeInvariant Control Design for a Pressurized Water Reactor

    E-Print Network [OSTI]

    Bodenheimer, Bobby

    . The plant can thus have widely varying dynamics over the operating range. The controllers designed perform to a description of the problem statement. Section 4 describes the identification and modelling of the plant. Se the worst­case time variation of a measurable parameter which enters the plant in a linear fractional manner

  20. CONTINUING AND PROFESSIONAL EDUCATION extension.ucdavis.edu/certificates

    E-Print Network [OSTI]

    Ferrara, Katherine W.

    and Construction Professionals Fall 2011 n Construction Management n Construction Estimating n Green Building and Sustainable Design O N LI N E n Sustainability and the Built Environment n Renewable Energy n Solar Energy.ucdavis.edu/business Green Building and Sustainable Design Certificate Program Stay on the forefront of the design, planning

  1. Forest management certification

    E-Print Network [OSTI]

    Liskiewicz, Maciej

    to the Montreal Process, certification systems lay out a set of criteria for sustainable management. Consumers. Each captures unique aspects of forestry and natural- resources management, offers indoor and outdoor.va.nrcs.usda.gov NRCS's natural resources conservation programs help people reduce soil erosion, enhance water supplies

  2. Database Management Certificate Program

    E-Print Network [OSTI]

    Rose, Michael R.

    Database Management Certificate Program DATABASE TECHNOLOGIES 13 Today, organizations in every to obtain general data analysis and management skills or apply both Oracle and Microsoft SQL Server database capabilities of relational databases in conjunction with the new "Big Data" management frameworks for data

  3. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

    2009-03-10T23:59:59.000Z

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  4. Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels (I-NERI Annual Report)

    SciTech Connect (OSTI)

    Petti, David Andrew; Maki, John Thomas; Languille, Alain; Martin, Philippe; Ballinger, Ronald

    2002-11-01T23:59:59.000Z

    The objective of this INERI project is to develop improved fuel behavior models for gas reactor coated particle fuels and to develop improved coated-particle fuel designs that can be used reliably at very high burnups and potentially in fast gas-cooled reactors. Thermomechanical, thermophysical, and physiochemical material properties data were compiled by both the US and the French and preliminary assessments conducted. Comparison between U.S. and European data revealed many similarities and a few important differences. In all cases, the data needed for accurate fuel performance modeling of coated particle fuel at high burnup were lacking. The development of the INEEL fuel performance model, PARFUME, continued from earlier efforts. The statistical model being used to simulate the detailed finite element calculations is being upgraded and improved to allow for changes in fuel design attributes (e.g. thickness of layers, dimensions of kernel) as well as changes in important material properties to increase the flexibility of the code. In addition, modeling of other potentially important failure modes such as debonding and asphericity was started. A paper on the status of the model was presented at the HTR-2002 meeting in Petten, Netherlands in April 2002, and a paper on the statistical method was submitted to the Journal of Nuclear Material in September 2002. Benchmarking of the model against Japanese and an older DRAGON irradiation are planned. Preliminary calculations of the stresses in a coated particle have been calculated by the CEA using the ATLAS finite element model. This model and the material properties and constitutive relationships will be incorporated into a more general software platform termed Pleiades. Pleiades will be able to analyze different fuel forms at different scales (from particle to fuel body) and also handle the statistical variability in coated particle fuel. Diffusion couple experiments to study Ag and Pd transport through SiC were conducted. Analysis and characterization of the samples continues. Two active transport mechanisms are proposed: diffusion in SiC and release through SiC cracks or another, as yet undetermined, path. Silver concentration profiles determined by XPS analysis suggest diffusion within the SiC layer, most likely dominated by grain boundary diffusion. However, diffusion coefficients calculated from mass loss measurements suggest a much faster release path, postulated as small cracks or flaws that provide open paths with little resistance to silver migration. Work is ongoing to identify and characterize this path. Work on Pd behavior has begun and will continue next year.

  5. Effects of Gadolinium and Europium on the Design and Submersion Criticality of a Fast Spectrum Space Reactor

    SciTech Connect (OSTI)

    King, Jeffrey C.; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies, The University of New Mexico, Albuquerque, NM, 87131 (United States); Chemical and Nuclear Engineering Department, The University of New Mexico, Albuquerque, NM, 87131 (United States)

    2005-02-06T23:59:59.000Z

    Gadolinium-155 and europium-151 are examined as alternative spectral shift absorbers to rhenium in the Scalable AMTEC Integrated Reactor space power System (SAIRS) heat-pipe reactor. Spectral shift absorbers counteract the reactivity increase when a compact, highly-enriched space nuclear reactor is submerged in seawater or wet sand and flooded following a launch abort accident. After all excess rhenium is removed from the reactor core, gadolinium-155 or europium-151 is added to the core in the form of a 0.1 mm oxide coating on the inside of the reactor vessel and/or as a nitride additive to the UN fuel. To compensate for increased parasitic neutron absorption, the UN fuel enrichment in the SAIRS reactor is increased to from 83.5% to a maximum of 94%. With 12 atom% 155GdN added to the reactor fuel, the outer diameter of the axial reflector decreased by 2 cm, and with a 155Gd2O3 coating on the inside of the reactor vessel, the reactor has $2.47 of excess reactivity at Beginning of Mission (compared to $2.08 for the rhenium base-case) and a worst case submersion and flooding accident reactivity of -$1.12 (compared to -$0.93 for the base-case). The resulting reactor and shield weigh 951.20 kg, for a savings of 100.94 kg over the base-case. When 9 atom% 151EuN is used in the fuel, the outer diameter of the axial reflector is reduced by 4 cm, and the reactor has $2.53 excess reactivity and -$1.13 of reactivity in the worst-case submersion and flooding accident scenario. The europium-case represents a mass savings of 143.16 kg over the base-case for a total reactor and shield mass of 908.98 kg.

  6. Significance of analog instrumentation - design philosophy of replacement dump arrest unit at Pickering Station Candu Reactor

    SciTech Connect (OSTI)

    Miller, J.F.; McDowell, R.W. [GAMMA-METRICS, San Diego, CA (United States)

    1996-12-31T23:59:59.000Z

    This paper discusses the differences of opinion concerning power plant instrumentation, including safety systems. One popular view point is that modem instrumentation must be microprocessor-based to be acceptable. An alternative view point is that properly designed analog instrumentation is recommended in some applications and has proven to be viable based upon performance and experience. A practical example is discussed in detail, explaining how a combination of discrete analog circuitry, combined with discrete digital circuitry provides a robust solution to a complex instrumentation replacement problem. In this application, a microprocessor-based instrument was designed as a replacement for an obsolete analog instrument. Due to severe licensing difficulties, the instrument was redesigned as a combination of discrete analog and digital circuitry. In the implementation of this circuitry, all complex testing functions of the improved microprocessor-based instrument were accommodated and system accuracy and performance were not compromised over the micro-processor-based instrument. The instrument has met all requirements for reliability and EMI/RFI susceptibility, as well as isolation of analog outputs and the ability to withstand severe transient noise on inputs and outputs without adversely affecting performance.

  7. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01T23:59:59.000Z

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  8. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01T23:59:59.000Z

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  9. Th/U-233 Multi-recycle in Pressurized Water Reactors: Feasibility Study of Multiple Homogeneous and Heterogeneous Assembly Designs

    E-Print Network [OSTI]

    Kemner, Ken

    Th/U-233 Multi-recycle in Pressurized Water Reactors: Feasibility Study of Multiple Homogeneous in the LWR fuel cycle. The possibility for thorium utilization in a multi-recycle system has also been fuel multi-recycle in current LWRs, focusing on pressurized water reactors (PWRs). Approaches

  10. Degree Information 37 Transcripted Graduate Certificate Programs

    E-Print Network [OSTI]

    program Energy Sustainability Engineering Certificate Program Graduate Certificate in Advanced the student's career goals. Energy Sustainability Engineering Certificate. The graduate level Energy--environmentally benign, economically competitive, and socially acceptable--to engineer the sustainability of energy

  11. A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...

    Office of Scientific and Technical Information (OSTI)

    A reactor and associated power plant designed to produce 1.05 Mwh and 3.535 Mwh of steam for heating purposes are described. The total thermal output of the reactor is 10 Mwh....

  12. Certification/enforcement analysis

    SciTech Connect (OSTI)

    None

    1980-06-01T23:59:59.000Z

    Industry compliance with minimum energy efficiency standards will be assured through a two-part program approach of certification and enforcement activities. The technical support document (TSD) presents the analyses upon which the proposed rule for assuring that consumer product comply with applicable energy efficiency standards is based. Much of the TSD is based upon support provided DOE by Vitro Laboratories. The OAO Corporation provided additional support in the development of the sampling plan incorporated in the proposed rule. Vitro's recommended approach to appliance certification and enforcement, developed after consideration of various program options, benefits, and impacts, establishes the C/E program framework, general criteria, and procedures for assuring a specified level of energy efficiency performance of covered consumer products. The results of the OAO analysis are given in Volume II of the TSD.

  13. 148 SDSU Curriculum Guide 2010 University Guidelines for Certificate Programs

    E-Print Network [OSTI]

    Ponce, V. Miguel

    148 SDSU Curriculum Guide 2010 University Guidelines for Certificate Programs (Guidelines Based on Executive Order 806) Types of Certificate Programs San Diego State University offers two types of certificate programs: · Academic certificate programs, and · Professional certificate programs. Academic

  14. TORT certification package

    SciTech Connect (OSTI)

    Frost, R.L.

    1993-10-01T23:59:59.000Z

    The TORT code has been certified. TORT is a three-dimensional discrete ordinates transport theory code, than can solve neutron, photon, or coupled neutron/photon problems. The code will be used primarily for shielding and radiation field calculations SRS. As defined in this work, certification dies not imply validation. The code must be validated for a particular type of calculation before it can be used for critical applications.

  15. Russian-Designed Reactors. Hearing before the Subcommittee on Energy and Power of the Committee on Energy and Commerce, House of Representatives, One Hundred Third Congress, First Session, October 28, 1993

    SciTech Connect (OSTI)

    Not Available

    1994-01-01T23:59:59.000Z

    This hearing focuses on the USA's effort to upgrade the safety of Soviety-designed nuclear power reactors including development of energy alternatives, and effective assistance, addressing the most serious problems. Testifying are Hazel O'Leary, US DOE; I. Selin, NRC, and Elizabeth Verville, coordinator of Reactor Safety and International Science and Technology Center, DOS.

  16. Central Characterization Program (CCP) Transuranic Waste Certification...

    Office of Environmental Management (EM)

    Certification Plan Central Characterization Program (CCP) Transuranic Waste Certification Plan This document was used to determine facts and conditions during the Department of...

  17. United Nations Conference on Trade and Development Making Certification Work for Sustainable

    E-Print Network [OSTI]

    United Nations Conference on Trade and Development Making Certification Work for Sustainable ..................................................................... 31 Chapter 5. Making certification work for sustainable development: the way ahead Development: The Case of Biofuels United Nations New York and Geneva, 2008 #12;ii Note The designations

  18. Conceptual design of a lead-bismuth cooled fast reactor with in-vessel direct-contact steam generation

    E-Print Network [OSTI]

    Buongiorno, Jacopo, 1971-

    2001-01-01T23:59:59.000Z

    The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid ...

  19. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

    K. T. Assessment of Candidate Molten Salt Coolants for theK. T. Assessment of Candidate Molten Salt Coolants for thebeginning efforts for a molten salt reactor (MSR) program.

  20. The design of a functionally graded composite for service in high temperature lead and lead-bismuth cooled nuclear reactors

    E-Print Network [OSTI]

    Short, Michael Philip

    2010-01-01T23:59:59.000Z

    A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700°C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required ...

  1. Thermal hydraulic design of a 2400 MW t?h? direct supercritical CO?-cooled fast reactor

    E-Print Network [OSTI]

    Pope, Michael A. (Michael Alexander)

    2006-01-01T23:59:59.000Z

    The gas cooled fast reactor (GFR) has received new attention as one of the basic concepts selected by the Generation-IV International Forum (GIF) for further investigation. Currently, the reference GFR is a helium-cooled ...

  2. Conceptual Design of a Lead-Bismuth Cooled Fast Reactor with In-Vessel Direct-Contact Steam Generation

    E-Print Network [OSTI]

    Buongiorno, J.

    The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid ...

  3. Application of USNRC NUREG/CR-6661 and draft DG-1108 to evolutionary and advanced reactor designs

    SciTech Connect (OSTI)

    Chang 'Apollo', Chen [Apollo Consulting, Inc., Surprise, AZ 85374-4605 (United States)

    2006-07-01T23:59:59.000Z

    For the seismic design of evolutionary and advanced nuclear reactor power plants, there are definite financial advantages in the application of USNRC NUREG/CR-6661 and draft Regulatory Guide DG-1108. NUREG/CR-6661, 'Benchmark Program for the Evaluation of Methods to Analyze Non-Classically Damped Coupled Systems', was by Brookhaven National Laboratory (BNL) for the USNRC, and Draft Regulatory Guide DG-1108 is the proposed revision to the current Regulatory Guide (RG) 1.92, Revision 1, 'Combining Modal Responses and Spatial Components in Seismic Response Analysis'. The draft Regulatory Guide DG-1108 is available at http://members.cox.net/apolloconsulting, which also provides a link to the USNRC ADAMS site to search for NUREG/CR-6661 in text file or image file. The draft Regulatory Guide DG-1108 removes unnecessary conservatism in the modal combinations for closely spaced modes in seismic response spectrum analysis. Its application will be very helpful in coupled seismic analysis for structures and heavy equipment to reduce seismic responses and in piping system seismic design. In the NUREG/CR-6661 benchmark program, which investigated coupled seismic analysis of structures and equipment or piping systems with different damping values, three of the four participants applied the complex mode solution method to handle different damping values for structures, equipment, and piping systems. The fourth participant applied the classical normal mode method with equivalent weighted damping values to handle differences in structural, equipment, and piping system damping values. Coupled analysis will reduce the equipment responses when equipment, or piping system and structure are in or close to resonance. However, this reduction in responses occurs only if the realistic DG-1108 modal response combination method is applied, because closely spaced modes will be produced when structure and equipment or piping systems are in or close to resonance. Otherwise, the conservatism in the current Regulatory Guide 1.92, Revision 1, will overshadow the advantage of coupled analysis. All four participants applied the realistic modal combination method of DG-1108. Consequently, more realistic and reduced responses were obtained. (authors)

  4. CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

    SciTech Connect (OSTI)

    Charak, I.; Kier, P.H. [Argonne National Lab., IL (United States)

    1995-04-01T23:59:59.000Z

    Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule ({section}50.62); (2) station blackout ({section}50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term ({section}50.34(f) and {section}100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements ({section}50.55a, etc); (6) ECCS acceptance criteria ({section}50.46)(b); (7) combustible gas control ({section}50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle ({section}51.51); and (11) (standards {section}50.55a).

  5. DORT certification package

    SciTech Connect (OSTI)

    Frost, R.L.

    1994-01-01T23:59:59.000Z

    The DORT code has been certified. DORT is a two-dimensional discrete ordinates transport theory code, that can solve neutron, photon, or coupled neutron/photon problems. It is anticipated that DORT will be used for criticality calculations as well as for shielding and radiation field analysis at SRS. In addition to the DORT module itself, 5 utility programs that are useful in certain DORT applications have been certified. These modules are: GIP, DOS, GRTUNCL, BNDRYS, and RTFLUM. As defined in this work, certification does not imply validation. These codes must be validated for a particular type of calculation before they can be used for critical applications.

  6. Certification des structures d'hospitalisation domicile VS certification ISO 9000

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Certification des structures d'hospitalisation ŕ domicile VS certification ISO 9000 RYM BEN comparative entre la certification ISO 9000 (International Organisation for Standardization) et la certification des HAD. Il s'agit de comparer la certification ISO 9000 ŕ celle des établissements de santé et

  7. Design of an Actinide Burning, Lead or Lead-Bismuth Cooled Reactor that Produces Low Cost Electricity FY-01 Annual Report, October 2001

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Herring, James Stephen; Loewen, Eric Paul; Smolik, Galen Richard; Weaver, Kevan Dean; Todreas, N.

    2001-10-01T23:59:59.000Z

    The purpose of this collaborative Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) Laboratory Directed Research and Development (LDRD) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, plant engineering, material compatibility studies, and coolant activation. The publications derived from work on this project (since project inception) are listed in Appendix A.

  8. Design of an Actinide Burning, Lead or Lead-Bismuth Cooled Reactor That Produces Low Cost Electricty - FY-02 Annual Report

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo

    2002-10-01T23:59:59.000Z

    The purpose of this collaborative Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) Laboratory Directed Research and Development (LDRD) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, plant engineering, material compatibility studies, and coolant activation. The publications derived from work on this project (since project inception) are listed in Appendix A. This is the third in a series of Annual Reports for this project, the others are also listed in Appendix A as FY-00 and FY-01 Annual Reports.

  9. Journey to Leadership Certificate Program Application

    Broader source: Energy.gov [DOE]

    Application packet for Journey to Leadership Certificate Program Pilot course, includes description.

  10. MCNP certification package

    SciTech Connect (OSTI)

    Trumble, E.F.

    1992-08-01T23:59:59.000Z

    In response to a Department of Energy (DOE) request, Westinghouse Savannah River Company committed to certify all computer codes used in critical calculations at the site. Since the Monte Carlo Neutron Photon Transport (MCNP) code will be used to perform critical analyses involving criticality and shielding, the code must be certified. Certification as applied to existing computer codes includes the verification and validation process, placing the code in configuration control, and establishing user qualification standards and training requirements. All software intended for use in critical calculations must be certified. This report is intended to fulfill the requirements for the certification of the MCNP code, version 4.2, built June 11, 1992, by J.H. Hightower on the SRS CRAY. This report does not release MCNP for use under production status for any application for which a MCNP validation document does not exist. These validation documents will describe the specific range of applicability, limitations on use, results and biases for a particular MCNP application.

  11. Reactor Analysis and Process Synthesis for a Class of Complex Sequential Reactions

    E-Print Network [OSTI]

    Maccabi, Omid A

    2012-01-01T23:59:59.000Z

    construction for isothermal reactor networks, Computers andA systematic generatioin of reactor designs. I. IsothermalOptimization of complex reactor networks. I. Isothermal

  12. MULTIPHASE REACTOR MODELING FOR ZINC CHLORIDE CATALYZED COAL LIQUEFACTION

    E-Print Network [OSTI]

    Joyce, Peter James

    2011-01-01T23:59:59.000Z

    Chemical Engineering University of California Berkeley~ California 94720 ABSTRACT A generalized reactor design

  13. Numerical simulations of epitaxial growth process in MOVPE reactor as a tool for design of modern semiconductors for high power electronics

    SciTech Connect (OSTI)

    Skibinski, Jakub; Wejrzanowski, Tomasz [Warsaw University of Technology, Faculty of Materials Science and Engineering, Woloska 141, 02507 Warsaw (Poland); Caban, Piotr [Institute of Electronic Materials Technology, Wolczynska 133, 01919 Warsaw (Poland); Kurzydlowski, Krzysztof J. [Warsaw University of Technology, Faculty of Materials Science and Engineering Woloska, 141, 02507 Warsaw (Poland)

    2014-10-06T23:59:59.000Z

    In the present study numerical simulations of epitaxial growth of gallium nitride in Metal Organic Vapor Phase Epitaxy reactor AIX-200/4RF-S is addressed. Epitaxial growth means crystal growth that progresses while inheriting the laminar structure and the orientation of substrate crystals. One of the technological problems is to obtain homogeneous growth rate over the main deposit area. Since there are many agents influencing reaction on crystal area such as temperature, pressure, gas flow or reactor geometry, it is difficult to design optimal process. According to the fact that it's impossible to determine experimentally the exact distribution of heat and mass transfer inside the reactor during crystal growth, modeling is the only solution to understand the process precisely. Numerical simulations allow to understand the epitaxial process by calculation of heat and mass transfer distribution during growth of gallium nitride. Including chemical reactions in numerical model allows to calculate the growth rate of the substrate and estimate the optimal process conditions for obtaining the most homogeneous product.

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

    SciTech Connect (OSTI)

    Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

    2013-11-29T23:59:59.000Z

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

  15. The Graduate Certificate in Emergency Management and Business Continuity

    E-Print Network [OSTI]

    Bieber, Michael

    company. A customized Graduate Certificate can also be offered to match your compa- ny's area such as intentional bombing or the result of a positive event such as an unexpected demand for a company's product, evaluation, and improvement WHAT COURSES ARE OFFERED? Select two from: IS 613 Design of Emergency Management

  16. Renewable Energy Certificate Program

    SciTech Connect (OSTI)

    Gwendolyn S. Andersen

    2012-07-17T23:59:59.000Z

    This project was primarily to develop and implement a curriculum which will train undergraduate and graduate students at the University seeking a degree as well as training for enrollees in a special certification program to prepare individuals to be employed in a broad range of occupations in the field of renewable energy and energy conservation. Curriculum development was by teams of Saint Francis University Faculty in the Business Administration and Science Departments and industry experts. Students seeking undergraduate and graduate degrees are able to enroll in courses offered within these departments which will combine theory and hands-on training in the various elements of wind power development. For example, the business department curriculum areas include economic modeling, finance, contracting, etc. The science areas include meteorology, energy conversion and projection, species identification, habitat protection, field data collection and analysis, etc.

  17. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01T23:59:59.000Z

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  18. State of Nevada comments on the OCRWM from-reactor spent fuel shipping cask preliminary design reports

    SciTech Connect (OSTI)

    Halstead, R.J.; Audin, L.; Hoskins, R.E.; Snedeker, D.F.

    1990-12-01T23:59:59.000Z

    The design of spent fuels shipping casks is described. Two casks from two different contractors are presented. The design needs are based on the OCRWM'S program specifications. (CBS)

  19. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

    the field of chemical engineering, design of scaled systemsDesign Approach to Safety and Reliability Yields Great Benefits. Chemical Engineering

  20. D-D tokamak reactor studies

    SciTech Connect (OSTI)

    Evans, K.E. Jr.; Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Finn, P.A.; Jung, J.; Mattas, R.F.; Misra, B.; Smith, D.L.; Stevens, H.C.

    1980-11-01T23:59:59.000Z

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated.

  1. Design of slurry bubble column reactors: novel technique for optimum catalyst size selection contractual origin of the invention

    DOE Patents [OSTI]

    Gamwo, Isaac K. (Murrysville, PA); Gidaspow, Dimitri (Northbrook, IL); Jung, Jonghwun (Naperville, IL)

    2009-11-17T23:59:59.000Z

    A method for determining optimum catalyst particle size for a gas-solid, liquid-solid, or gas-liquid-solid fluidized bed reactor such as a slurry bubble column reactor (SBCR) for converting synthesis gas into liquid fuels considers the complete granular temperature balance based on the kinetic theory of granular flow, the effect of a volumetric mass transfer coefficient between the liquid and the gas, and the water gas shift reaction. The granular temperature of the catalyst particles representing the kinetic energy of the catalyst particles is measured and the volumetric mass transfer coefficient between the gas and liquid phases is calculated using the granular temperature. Catalyst particle size is varied from 20 .mu.m to 120 .mu.m and a maximum mass transfer coefficient corresponding to optimum liquid hydrocarbon fuel production is determined. Optimum catalyst particle size for maximum methanol production in a SBCR was determined to be in the range of 60-70 .mu.m.

  2. Framework for Certification of Fish Propagation, Protection and Monitoring Facilities. Final Report.

    SciTech Connect (OSTI)

    Carlson, Thomas J.; Costello, Ronald J.

    1997-06-01T23:59:59.000Z

    A conceptual framework for certification of fish production and monitoring facilities including software templates to expedite implementation of the framework are presented. The framework is based on well established and widely utilized project management techniques. The implementation templates are overlays for Microsoft Professional Office software products: Excel, Word, and Project. Use of the software templates requires Microsoft Professional Office. The certification framework integrates two classical project management processes with a third process for facility certification. These processes are: (1) organization and definition of the project, (2) acquisition and organization of project documentation, and (3) facility certification. The certification process consists of systematic review of the production processes and the characteristics of the produced product. The criteria for certification review are the plans and specifications for the products and production processes that guided development of the facility. The facility is certified when the production processes are operating as designed and the product produced meets specifications. Within this framework, certification is a performance based process, not dissimilar from that practiced in many professions and required for many process, or a product meets professional/industry standards of performance. In the case of fish production facilities, the certifying authority may be diffuse, consisting of many entities acting through a process such as NEPA. A cornerstone of certification is accountability, over the long term, for the operation and products of a facility. This is particularly important for fish production facilities where the overall goal of the facility may require decades to accomplish.

  3. ardennes reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    d'un parasite cycle Paris-Sud XI, Universit de 12 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  4. aprf reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    of the specimen. The sensitivity Paris-Sud XI, Universit de 6 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  5. a-2 reactor bohunice: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Barbara Ricci; Virginia Strati; Gerti Xhixha 2014-11-24 7 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  6. a-1 reactor bohunice: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    of degree one rather than rational points. Nguyen Le Dang Thi 7 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  7. State of Nevada comments on the OCRWM from-reactor spent fuel shipping cask preliminary design reports

    SciTech Connect (OSTI)

    Halstead, R.J.; Audin, L.; Hoskins, R.E.; Snedeker, D.F.

    1990-12-01T23:59:59.000Z

    The design of spent fuels shipping casks is described. Two casks from two different contractors are presented. The design needs are based on the OCRWM`S program specifications. (CBS)

  8. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

    reprocessing to recover fissionable material, FHR fuel handling systems must be designed to facilitate the application of IAEA safeguards.

  9. Examples of the use of PSA in the design process and to support modifications at two research reactors

    SciTech Connect (OSTI)

    Johnson, D.H.; Bley, D.C.; Lin, J.C. [PLG, Inc., Newport Beach, CA (United States); Ramsey, C.T.; Linn, M.A. [Oak Ridge National Lab., TN (United States)

    1994-03-01T23:59:59.000Z

    Many, if not most, of the world`s commercial nuclear power plants have been the subject of plant-specific probabilistic safety assessments (PSA). A growing number of other nuclear facilities as well as other types of industrial installations have been the focus of plant-specific PSAs. Such studies have provided valuable information concerning the nature of the risk of the individual facility and have been used to identify opportunities to manage that risk. This paper explores the risk management activities associated with two research reactors in the United States as a demonstration of the versatility of the use of PSA to support risk-related decision making.

  10. Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Ellis, Ronald James [ORNL; Gehin, Jess C [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL

    2007-11-01T23:59:59.000Z

    This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology, several engineering proof-of-principle tests would be required. The RERTR program is currently conducting a series of generic fuel qualification tests at the Advanced Test Reactor. A review of these tests and a review of the safety basis for the current, HEU fuel cycle led to the identification of a set of HFIR-specific fuel qualification tests. Much additional study is required to formulate a HFIR-specific fuel qualification plan from this set. However, one such test - creating a graded fuel profile across a flat foil - has been initiated with promising results.

  11. PREPARING FOR A SUCCESSFUL EVMS CERTIFICATION

    SciTech Connect (OSTI)

    CROWE SL; BASCHE AD

    2011-02-09T23:59:59.000Z

    The client, a government agency, requires its contractor to obtain an Earned Value Management System (EVMS) certification that meets the intent of ANSI/EIA-748-B, Earned Value Management Systems. The contractor has extensive experience with certification preparation, having completed two certifications within two years. Information from a previous EVMS certification and internal system surveillances are used to prepare for client-based EVMS certifications and bi-annual surveillances. The contractor also sent members of its group to assist other companies preparing for surveillances and certifications to perform 'Black Hat Reviews.' This paper is a lessons learned on preparing a team for EVMS certification. The information is also applicable for surveillances, since the contractor prepares its team for the surveillance in the same manner as the initial certification. Some of the areas covered include required documents, tracing the data through the systems, Control Account Manager (CAM) preparation, and system verification.

  12. STATE OF CALIFORNIA REFRIGERATED WAREHOUSE INSTALLATION CERTIFICATE

    E-Print Network [OSTI]

    STATE OF CALIFORNIA REFRIGERATED WAREHOUSE INSTALLATION CERTIFICATE CEC-RWH-INST (Revised 08 # BUILDING TYPE Refrigerated Warehouse PHASE OF CONSTRUCTION New Construction Addition Alteration If more By the Enforcement Agency #12;STATE OF CALIFORNIA REFRIGERATED WAREHOUSE INSTALLATION CERTIFICATE CEC

  13. Policy -Online Business Management Certificate Program

    E-Print Network [OSTI]

    Hanson, Stephen José

    Advisor-Instructor's assignments. Refund Policy (for Online Certificate program only - Lean Six Sigma programs. Grievance Policy If you are a participant in our Online Certificate Program or Lean/Six Sigma

  14. Community Leadership Certificate Program Program Overview

    E-Print Network [OSTI]

    Saldin, Dilano

    Community Leadership Certificate Program Program Overview The undergraduate Community Leadership Certificate Program is an interdisciplinary program where students from any major can explore leadership in community engagement and leadership. The program is based on a social justice model of leadership

  15. Computational Science and Engineering Certification for Chemistry

    E-Print Network [OSTI]

    Gilbert, Matthew

    Computational Science and Engineering Certification for Chemistry The Computational Science that are distinct from already- required coursework. To receive a certificate in "Computational Science a solid base in problem solving using computation as a major tool for modeling complicated problems

  16. Design and Certification of Industrial Predictive Controllers

    E-Print Network [OSTI]

    Dutta, Abhishek

    2014-09-24T23:59:59.000Z

    Procedure . . . . . . . . . . . . . . 4-6 4.4 Test case: Non-collocated mass-spring-damper . . . . . . . . . . 4-8 4.4.1 Mass-spring-damper setup . . . . . . . . . . . . . . . . . 4-8 4.4.2 PID control . . . . . . . . . . . . . . . . . . . . . . . . . 4-10 4... by distributed NMPC and (b): PID control. . . . . . . . . . . . . . . . . . . . . 6-15 6.7 (a): Learning the true motor damping coefficients by the distributed RLS method and (b): Robust performance of the distributed NMPC after learning the correct damping...

  17. T-706: Microsoft Fraudulent Digital Certificate Issued by DigiNotar

    Broader source: Energy.gov [DOE]

    Fraudulent Digital Certificate Issued by DigiNotar, a Certification Authority Present in the Trusted Root Certification Authorities

  18. GRISET certification package

    SciTech Connect (OSTI)

    Taylor, J.J.; Cefus, G.R.

    1992-11-01T23:59:59.000Z

    The GRISET (GRIMHX setup) computer code is an input driver for the three dimensional G computer code (Reference 4). GRISET provides a method for automating the creation of G geometry and associated problem specific information such as the reactor maps and the cross section data references. GRISET provides methods which allow the user to model control rod positioning, safety rod insertion with any number failing, control rod withdrawal or insertion, and full or partial melting of control rods. GRISET also calculates the radial shape factor, tilt, roof-top-ratios, and the assembly power peaking.

  19. ARTS LEADERSHIP Graduate Academic Certificate Program

    E-Print Network [OSTI]

    Mohanty, Saraju P.

    ARTS LEADERSHIP Graduate Academic Certificate Program The Certificate in Arts Leadership leadership. Within a broad definition of leadership as "a process whereby an individual influences a group of individuals to achieve a common goal," the Certificate in Arts Leadership addresses four essential components

  20. Division of Human Resources Payroll Certification Process

    E-Print Network [OSTI]

    Meyers, Steven D.

    Division of Human Resources Payroll Certification Process Questions (813) 974-7955 Payroll/Payroll Processing Rev. 04/2010 Payroll Certification is the process by which departments submit to Payroll the hours to be paid for each employee within each department. This process utilizes the online Certification System

  1. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1994-12-31T23:59:59.000Z

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future.

  2. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03T23:59:59.000Z

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  3. Microchannel Reactor System Design & Demonstration For On-Site H2O2 Production by Controlled H2/O2 Reaction

    SciTech Connect (OSTI)

    Adeniyi Lawal

    2008-12-09T23:59:59.000Z

    We successfully demonstrated an innovative hydrogen peroxide (H2O2) production concept which involved the development of flame- and explosion-resistant microchannel reactor system for energy efficient, cost-saving, on-site H2O2 production. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for controlled direct combination of H2 and O2 in all proportions including explosive regime, at a low pressure and a low temperature to produce about 1.5 wt% H2O2 as proposed. In the second phase of the program, as a prelude to full-scale commercialization, we demonstrated our H2O2 production approach by ‘numbering up’ the channels in a multi-channel microreactor-based pilot plant to produce 1 kg/h of H2O2 at 1.5 wt% as demanded by end-users of the developed technology. To our knowledge, we are the first group to accomplish this significant milestone. We identified the reaction pathways that comprise the process, and implemented rigorous mechanistic kinetic studies to obtain the kinetics of the three main dominant reactions. We are not aware of any such comprehensive kinetic studies for the direct combination process, either in a microreactor or any other reactor system. We showed that the mass transfer parameter in our microreactor system is several orders of magnitude higher than what obtains in the macroreactor, attesting to the superior performance of microreactor. A one-dimensional reactor model incorporating the kinetics information enabled us to clarify certain important aspects of the chemistry of the direct combination process as detailed in section 5 of this report. Also, through mathematical modeling and simulation using sophisticated and robust commercial software packages, we were able to elucidate the hydrodynamics of the complex multiphase flows that take place in the microchannel. In conjunction with the kinetics information, we were able to validate the experimental data. If fully implemented across the whole industry as a result of our technology demonstration, our production concept is expected to save >5 trillion Btu/year of steam usage and >3 trillion Btu/year in electric power consumption. Our analysis also indicates >50 % reduction in waste disposal cost and ~10% reduction in feedstock energy. These savings translate to ~30% reduction in overall production and transportation costs for the $1B annual H2O2 market.

  4. Directory of Certificates of Compliance for radioactive materials packages: Certificates of Compliance. Volume 2, Revision 18

    SciTech Connect (OSTI)

    NONE

    1995-10-01T23:59:59.000Z

    The purpose of this directory is to make available a convenient source of information on packagings which have been approved by the US Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance Number is included at the front of Volumes 1 and 2. An alphabetical listing by user name is included in the back of Volume 3 of approved QA programs. The reports include a listing of all users of each package design and approved QA programs prior to the publication date.

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

    SciTech Connect (OSTI)

    Criddle, Craig S.; Wu, Weimin

    2013-04-17T23:59:59.000Z

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

  6. Microfluidic reactors for the synthesis of nanocrystals

    E-Print Network [OSTI]

    Yen, Brian K. H

    2007-01-01T23:59:59.000Z

    Several microfluidic reactors were designed and applied to the synthesis of colloidal semiconductor nanocrystals (NCs). Initially, a simple single-phase capillary reactor was used for the synthesis of CdSe NCs. Precursors ...

  7. Plant Design and Cost Estimation of a Natural Circulation Lead-Bismuth Reactor with Helium Power Conversion Cycle

    E-Print Network [OSTI]

    Kim, D.

    The analysis of an indirect helium power conversion system with lead-bismuth natural circulation primary system has been performed. The work of this report is focused on 1) identifying the allowable design region for the ...

  8. Plant Design and Cost Assessment of Forced Circulation Lead-Bismuth Cooled Reactor with Conventional Power Conversion Cycles

    E-Print Network [OSTI]

    Dostal, Vaclav

    Cost of electricity is the key factor that determines competitiveness of a power plant. Thus the proper selection, design and optimization of the electric power generating cycle is of main importance. This report makes an ...

  9. Design of an experimental loop for post-LOCA heat transfer regimes in a Gas-cooled Fast Reactor

    E-Print Network [OSTI]

    Cochran, Peter A. (Peter Andrew)

    2005-01-01T23:59:59.000Z

    The goal of this thesis is to design an experimental thermal-hydraulic loop capable of generating accurate, reliable data in various convection heat transfer regimes for use in the formulation of a comprehensive convection ...

  10. Plant Design and Cost Estimation of a Natural Circulation Lead-Bismuth Reactor with Steam Power Conversion Cycle

    E-Print Network [OSTI]

    Kim, D.

    The analysis of an indirect steam power conversion system with lead-bismuth natural circulation primary system has been performed. The work of this report is focused on 1) identifying the allowable design region for the ...

  11. Conceptual design study FY 1981: synfuels from fusion - using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

    SciTech Connect (OSTI)

    Krikorian, O.H. (ed.)

    1982-02-09T23:59:59.000Z

    This report represents the second year's effort of a scoping and conceptual design study being conducted for the express purpose of evaluating the engineering potential of producing hydrogen by thermochemical cycles using a tandem mirror fusion driver. The hydrogen thus produced may then be used as a feedstock to produce fuels such as methane, methanol, or gasoline. The main objective of this second year's study has been to obtain some approximate cost figures for hydrogen production through a conceptual design study.

  12. Effect of improved target designs on the sup 238 Pu production at the Fast Flux Test Reactor

    SciTech Connect (OSTI)

    Karnesky, R.A.; Wootan, D.W.; Jordheim, D.P. (Westinghouse Hanford Co., Richland, WA (United States))

    1991-11-01T23:59:59.000Z

    This paper present the results of a series of calculations made to determine the {sup 238}Pu production potential of several advanced target assembly designs in the Fast Flux Test Facility (FFTF). These calculations show that by using advanced target designs the intimately mix the {sup 237}Np target material with an yttrium hydride moderator, the FFTF has the potential of producing up to 30 kg of high-quality {sup 238}Pu per year.

  13. FINANCIAL RISK MANAGER (FRM ) CERTIFICATION

    E-Print Network [OSTI]

    Chaudhuri, Sanjay

    Dec 2013 (Save 30%) (Save 15%) Module 1 Foundation of Risk Management & Quantitative Analysis 1086FINANCIAL RISK MANAGER (FRM® ) CERTIFICATION TRAINING PROGRAM 2014 JANUARY INTAKE 4 January to 10 1321 1552 Module 2 Financial Markets and Products & Valuation and Risk Models 1637 1985 2333 Module 3

  14. Postgraduate Certificate in Geothermal Energy

    E-Print Network [OSTI]

    Auckland, University of

    Postgraduate Certificate in Geothermal Energy Technology The University of Auckland The University for development of geothermal fields is large and many countries are seeking to move away from fossil fuel power generation for both economic and environmental reasons. Global revenues for geothermal power were estimated

  15. Light Water Reactor Sustainability Program Operator Performance Metrics for Control Room Modernization: A Practical Guide for Early Design Evaluation

    SciTech Connect (OSTI)

    Ronald Boring; Roger Lew; Thomas Ulrich; Jeffrey Joe

    2014-03-01T23:59:59.000Z

    As control rooms are modernized with new digital systems at nuclear power plants, it is necessary to evaluate the operator performance using these systems as part of a verification and validation process. There are no standard, predefined metrics available for assessing what is satisfactory operator interaction with new systems, especially during the early design stages of a new system. This report identifies the process and metrics for evaluating human system interfaces as part of control room modernization. The report includes background information on design and evaluation, a thorough discussion of human performance measures, and a practical example of how the process and metrics have been used as part of a turbine control system upgrade during the formative stages of design. The process and metrics are geared toward generalizability to other applications and serve as a template for utilities undertaking their own control room modernization activities.

  16. Reactor Neutrino Experiments

    E-Print Network [OSTI]

    Jun Cao

    2007-12-06T23:59:59.000Z

    Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measurement are also briefed.

  17. Official Certificate List Level(s)* Academic Administrative

    E-Print Network [OSTI]

    Risk, Uncertainty and Decision Analysis EGRU335 Certificate in Engineering Thermal Energy Systems EGRU International Certificate ALSU198 Certificate for Biology in Engineering for Engineering Majors EGRU135 Level IESG310 Certificate in Engineering for Energy Sustainability EGRU340 Certificate in Engineering

  18. Design modification for the modular helium reactor for higher temperature operation and reliability studies for nuclear hydrogen production processes 

    E-Print Network [OSTI]

    Reza, S.M. Mohsin

    2009-05-15T23:59:59.000Z

    of graphite removed from the PSR to create this inlet path. With the removal of ~10% of the graphite from PSR the PVT is reduced from 541 0C to 421 0C. A new design for the graphite block core has been evaluated and optimized to reduce the inlet coolant...

  19. NEW OPTIMIZATION-BASED APPROACH TO CHEMICAL REACTOR SYNTHESIS TOWARDS THE FULL INTEGRATION OF REACTOR

    E-Print Network [OSTI]

    Van den Hof, Paul

    NEW OPTIMIZATION-BASED APPROACH TO CHEMICAL REACTOR SYNTHESIS ­ TOWARDS THE FULL INTEGRATION OF REACTOR DESIGN, OPERATION AND CONTROL Nikola Nikacevi1 , Adrie Huesman1 , Paul Van den Hof1 , Andrzej, Leeghwaterstraat 44, 2628 CA Delft, The Netherlands Conceptual chemical reactor design still follows traditional

  20. Optimization of the pyrolysis process of empty fruit bunch (EFB) in a fixed-bed reactor through a central composite design (CCD)

    SciTech Connect (OSTI)

    Mohamed, Alina Rahayu; Hamzah, Zainab; Daud, Mohamed Zulkali Mohamed [School of Bioprocess Engineering, Jejawi Complex of Academics (3), UniMAP, 02600 Arau Perlis (Malaysia)

    2014-07-10T23:59:59.000Z

    The production of crude palm oil from the processing of palm fresh fruit bunches in the palm oil mills in Malaysia hs resulted in a huge quantity of empty fruit bunch (EFB) accumulated. The EFB was used as a feedstock in the pyrolysis process using a fixed-bed reactor in the present study. The optimization of process parameters such as pyrolysis temperature (factor A), biomass particle size (factor B) and holding time (factor C) were investigated through Central Composite Design (CCD) using Stat-Ease Design Expert software version 7 with bio-oil yield considered as the response. Twenty experimental runs were conducted. The results were completely analyzed by Analysis of Variance (ANOVA). The model was statistically significant. All factors studied were significant with p-values < 0.05. The pyrolysis temperature (factor A) was considered as the most significant parameter because its F-value of 116.29 was the highest. The value of R{sup 2} was 0.9564 which indicated that the selected factors and its levels showed high correlation to the production of bio-oil from EFB pyrolysis process. A quadratic model equation was developed and employed to predict the highest theoretical bio-oil yield. The maximum bio-oil yield of 46.2 % was achieved at pyrolysis temperature of 442.15 °C using the EFB particle size of 866 ?m which corresponded to the EFB particle size in the range of 710–1000 ?m and holding time of 483 seconds.

  1. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2006

    SciTech Connect (OSTI)

    Primm, R. T. [ORNL] [ORNL; Ellis, R. J. [ORNL] [ORNL; Gehin, J. C. [ORNL] [ORNL; Clarno, K. T. [ORNL] [ORNL; Williams, K. A. [ORNL] [ORNL; Moses, D. L. [ORNL] [ORNL

    2006-11-01T23:59:59.000Z

    Neutronics and thermal-hydraulics studies show that, for equivalent operating power [85 MW(t)], a low-enriched uranium (LEU) fuel cycle based on uranium-10 wt % molybdenum (U-10Mo) metal foil with radially, “continuously graded” fuel meat thickness results in a 15% reduction in peak thermal flux in the beryllium reflector of the High Flux Isotope Reactor (HFIR) as compared to the current highly enriched uranium (HEU) cycle. The uranium-235 content of the LEU core is almost twice the amount of the HEU core when the length of the fuel cycle is kept the same for both fuels. Because the uranium-238 content of an LEU core is a factor of 4 greater than the uranium-235 content, the LEU HFIR core would weigh 30% more than the HEU core. A minimum U-10Mo foil thickness of 84 ?m is required to compensate for power peaking in the LEU core although this value could be increased significantly without much penalty. The maximum U-10Mo foil thickness is 457?m. Annual plutonium production from fueling the HFIR with LEU is predicted to be 2 kg. For dispersion fuels, the operating power for HFIR would be reduced considerably below 85 MW due to thermal considerations and due to the requirement of a 26-d fuel cycle. If an acceptable fuel can be developed, it is estimated that $140 M would be required to implement the conversion of the HFIR site at Oak Ridge National Laboratory from an HEU fuel cycle to an LEU fuel cycle. To complete the conversion by fiscal year 2014 would require that all fuel development and qualification be completed by the end of fiscal year 2009. Technological development areas that could increase the operating power of HFIR are identified as areas for study in the future.

  2. Cost/performance comparison between pulse columns and centrifugal contactors designed to process Clinch River Breeder Reactor fuel

    SciTech Connect (OSTI)

    Ciucci, J.A. Jr.

    1983-12-01T23:59:59.000Z

    A comparison between pulse columns and centrifugal contactors was made to determine which type of equipment was more advantageous for use in the primary decontamination cycle of a remotely operated fuel reprocessing plant. Clinch River Breeder Reactor (CRBR) fuel was chosen as the fuel to be processed in the proposed 1 metric tonne/day reprocessing facility. The pulse columns and centrifugal contactors were compared on a performance and total cost basis. From this comparison, either the pulse columns or the centrifugal contactors will be recommended for use in a fuel reprocessing plant built to reprocess CRBR fuel. The reliability, solvent exposure to radiation, required time to reach steady state, and the total costs were the primary areas of concern for the comparison. The pulse column units were determined to be more reliable than the centrifugal contactors. When a centrifugal contactor motor fails, it can be remotely changed in less than one eight hour shift. Pulse columns expose the solvent to approximately five times as much radiation dose as the centrifugal contactor units; however, the proposed solvent recovery system adequately cleans the solvent for either case. The time required for pulse columns to reach steady state is many times longer than the time required for centrifugal contactors to reach steady state. The cost comparison between the two types of contacting equipment resulted in centrifugal contactors costing 85% of the total cost of pulse columns when the contactors were stacked on three levels in the module. If the centrifugal contactors were all positioned on the top level of a module with the unoccupied volume in the module occupied by other equipment, the centrifugal contactors cost is 66% of the total cost of pulse columns. Based on these results, centrifugal contactors are recommended for use in a remotely operated reprocessing plant built to reprocess CRBR fuel.

  3. Remedial Action Certification Docket - Sodium Reactor Experiment (SRE)

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradleyTableSelling7 AugustAFRICAN3u ;;;:: A' 3 ct' RIDGE

  4. Advanced Safeguards Approaches for New Fast Reactors

    SciTech Connect (OSTI)

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

    2007-12-15T23:59:59.000Z

    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.

  5. Ssessment methodology for proliferation resistant fast breeder reactor

    E-Print Network [OSTI]

    Singh, Mohit, S.M. Massachusetts Institute of Technology

    2014-01-01T23:59:59.000Z

    Due to perceived proliferation risks, current US fast reactor designs have avoided the use of uranium blankets. While reducing the amount of plutonium produced, this omission also restrains the reactor design space and has ...

  6. UAF Green Department Certification UAF OFFICE OF SUSTAINABILITY

    E-Print Network [OSTI]

    Hartman, Chris

    /Documentation: Computers are set to energy saving modes. Description/Documentation: GREEN DEPARTMENT PLEDGE UAF Green Department Certification UAF OFFICE OF SUSTAINABILITY GREEN DEPARTMENT CERTIFICATION INSTRUCTIONS & FORMS PART I: GETTING STARTED & REQUIRED ACTIONS #12; UAF Green Department Certification

  7. Security Certification & Accreditation of Federal Information Systems A Tutorial

    E-Print Network [OSTI]

    Madisetti, Vijay K.

    Security Certification & Accreditation of Federal Information Systems A Tutorial An Introduction Madisetti, 06/29/2009 Security Certification & Assurance of Federal Information Systems Tutorial Tutorial Outline Objectives & Introduction: C&A Information Security Certification & Accreditation Foundations (as

  8. Tivoli Professional Certification e-Newsletter Check Out the Latest Tivoli Certifications and News

    E-Print Network [OSTI]

    & Certification Retirement -> Fund. of Enterprise Solutions Using IBM Tivoli Security 2007 Tivoli Storage Manager of the Month -> Shayne Gardener Promotions - Discounted Exams and Assessments -> "Early Bird" Specials & CERTIFICATION RETIREMENT For the latest news in exam withdrawals and certification retirement, go to: http

  9. Tivoli Professional Certification e-Check Out the Latest Tivoli Certifications

    E-Print Network [OSTI]

    Workshops -> IBM Tivoli Storage Manager V6.1 Administration Exam Withdrawal & Certification Retirement Rodriguez Promotions - Discounted Exams and Assessments -> "Early Bird" Specials for Newly Published Exams & CERTIFICATION RETIREMENT For the latest news in exam withdrawals and certification retirement, go to: http

  10. Motor Vehicle Rental Exemption Certificate THIS EXEMPTION CERTIFICATE IS NOT VALID FOR TAX-FREE REGISTRATION.

    E-Print Network [OSTI]

    Behmer, Spencer T.

    Motor Vehicle Rental Exemption Certificate THIS EXEMPTION CERTIFICATE IS NOT VALID FOR TAX-FREE REGISTRATION. THIS EXEMPTION CERTIFICATE MUST BE ATTACHED TO THE RENTAL CONTRACT. Make of Vehicle Motor or Vehicle Identification Number Year Model Body Style License Number The undersigned claims exemption from

  11. Chemical Reactor Analysis and Optimal Digestion

    E-Print Network [OSTI]

    Jumars, Pete

    J 310 Chemical Reactor Analysis and Optimal Digestion An optimal digestion theory can be readily derived from basic principles o f chemical reactor analysis and design Deborah L. Penry and Peter a reactor and an operating strategy that maximize the yield or yield rate of desired reaction products

  12. Personnel Selection, Training, Qualification, and Certification...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    6.2 Admin Chg 1, Personnel Selection, Training, Qualification, and Certification Requirements for DOE Nuclear Facilities by Sam Rosenbloom Functional areas: Administrative Change,...

  13. Personnel Selection, Training, Qualification, and Certification...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    6.2, Personnel Selection, Training, Qualification, and Certification Requirements for DOE Nuclear Facilities by Sam Rosenbloom Functional areas: Safety, The order establishes...

  14. FEMA - National Flood Insurance Program Elevation Certificate...

    Open Energy Info (EERE)

    and Instructions Jump to: navigation, search OpenEI Reference LibraryAdd to library Web Site: FEMA - National Flood Insurance Program Elevation Certificate and Instructions...

  15. Vermont Water Quality Certification Application for Hydroelectric...

    Open Energy Info (EERE)

    Vermont Water Quality Certification Application for Hydroelectric Facilities Jump to: navigation, search OpenEI Reference LibraryAdd to library Form: Vermont Water Quality...

  16. Renewable Energy Specifications, Testing and Certification Terms...

    Open Energy Info (EERE)

    Terms of Reference Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Renewable Energy Specifications, Testing and Certification Terms of Reference Agency...

  17. Arizona Corporation Commission Application for a Certificate...

    Open Energy Info (EERE)

    Application for a Certificate of Convenience and Necessity for Competitive Retail Electricity Services Jump to: navigation, search OpenEI Reference LibraryAdd to library Legal...

  18. CRADA ORGANIZATIONAL CONFLICTS OF INTEREST (OCI) CERTIFICATION

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    CRADA No. CRADA ORGANIZATIONAL CONFLICTS OF INTEREST (OCI) CERTIFICATION In accordance with Clause I.105 - Technology Transfer Mission, (n)(5)i of Contract No....

  19. Interfacial effects in fast reactors

    E-Print Network [OSTI]

    Saidi, Mohammad Said

    1979-01-01T23:59:59.000Z

    The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed ...

  20. National Certification Standard for Ground Source Heat Pump Personnel

    SciTech Connect (OSTI)

    Kelly, John [Geothermal Heat Pump Consortium] [Geothermal Heat Pump Consortium

    2013-07-31T23:59:59.000Z

    The National Certification Standard for the Geothermal Heat Pump Industry adds to the understanding of the barriers to rapid growth of the geothermal heat pump (GHP) industry by bringing together for the first time an analysis of the roles and responsibilities of each of the individual job tasks involved in the design and installation of GHP systems. The standard addresses applicable qualifications for all primary personnel involved in the design, installation, commissioning, operation and maintenance of GHP systems, including their knowledge, skills and abilities. The resulting standard serves as a foundation for subsequent development of curriculum, training and certification programs, which are not included in the scope of this project, but are briefly addressed in the standard to describe ways in which the standard developed in this project may form a foundation to support further progress in accomplishing those other efforts. Follow-on efforts may use the standard developed in this project to improve the technical effectiveness and economic feasibility of curriculum development and training programs for GHP industry personnel, by providing a more complete and objective assessment of the individual job tasks necessary for successful implementation of GHP systems. When incorporated into future certification programs for GHP personnel, the standard will facilitate increased consumer confidence in GHP technology, reduce the potential for improperly installed GHP systems, and assure GHP system quality and performance, all of which benefit the public through improved energy efficiency and mitigated environmental impacts of the heating and cooling of homes and businesses.

  1. POLICY GUIDANCE MEMORANDUM #27 Certification Program on Prohibited...

    Office of Environmental Management (EM)

    7 Certification Program on Prohibited Personnel Practices and Whistleblower Protections POLICY GUIDANCE MEMORANDUM 27 Certification Program on Prohibited Personnel Practices and...

  2. Vehicle Certification Test Fuel and Ethanol Flex Fuel Quality...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Vehicle Certification Test Fuel and Ethanol Flex Fuel Quality Vehicle Certification Test Fuel and Ethanol Flex Fuel Quality Breakout Session 2: Frontiers and Horizons Session 2-B:...

  3. Iowa Water and Wastewater Operators Seek SEP Certification in...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Iowa Water and Wastewater Operators Seek SEP Certification in New Pilot Program Iowa Water and Wastewater Operators Seek SEP Certification in New Pilot Program September 18, 2014 -...

  4. Volvo: Certification of an Integrated ISO 14001 and ISO 50001...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Assistance Superior Energy Performance Volvo: Certification of an Integrated ISO 14001 and ISO 50001SEP System Volvo: Certification of an Integrated ISO 14001 and ISO...

  5. Green Building Certification Systems Requirement for New Federal...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Green Building Certification Systems Requirement for New Federal Buildings and Major Renovations of Federal Buildings OIRA Comparison Document Green Building Certification Systems...

  6. Simulated Performance of the Integrated Passive Neutron Albedo Reactivity and Self-Interrogation Neutron Resonance Densitometry Detector Designed for Spent Fuel Measurement at the Fugen Reactor in Japan

    SciTech Connect (OSTI)

    Ulrich, Timothy J. II [Los Alamos National Laboratory; Lafleur, Adrienne M. [Los Alamos National Laboratory; Menlove, Howard O. [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory; Seya, Michio [Los Alamos National Laboratory; Bolind, Alan M. [Los Alamos National Laboratory

    2012-07-16T23:59:59.000Z

    An integrated nondestructive assay instrument, which combined the Passive Neutron Albedo Reactivity (PNAR) and the Self-Interrogation Neutron Resonance Densitometry (SINRD) techniques, is the research focus for a collaborative effort between Los Alamos National Laboratory (LANL) and the Japanese Atomic Energy Agency as part of the Next Generation Safeguard Initiative. We will quantify the anticipated performance of this experimental system in two physical environments: (1) At LANL we will measure fresh Low Enriched Uranium (LEU) assemblies for which the average enrichment can be varied from 0.2% to 3.2% and for which Gd laced rods will be included. (2) At Fugen we will measure spent Mixed Oxide (MOX-B) and LEU spent fuel assemblies from the heavy water moderated Fugen reactor. The MOX-B assemblies will vary in burnup from {approx}3 GWd/tHM to {approx}20 GWd/tHM while the LEU assemblies ({approx}1.9% initial enrichment) will vary from {approx}2 GWd/tHM to {approx}7 GWd/tHM. The estimated count rates will be calculated using MCNPX. These preliminary results will help the finalization of the hardware design and also serve a guide for the experiment. The hardware of the detector is expected to be fabricated in 2012 with measurements expected to take place in 2012 and 2013. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

  7. Development of Design and Simulation Tool for Hybrid Geothermal...

    Open Energy Info (EERE)

    performance analysis for various purposes, including LEED certification and optimization of building envelop andor HVAC system designs. In 2006, as a cooperative effort...

  8. Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

    SciTech Connect (OSTI)

    Shannon M. Bragg-Sitton

    2013-09-01T23:59:59.000Z

    Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

  9. Perspectives on reactor safety

    SciTech Connect (OSTI)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01T23:59:59.000Z

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

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

    SciTech Connect (OSTI)

    Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil)] [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil)] [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina)] [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)] [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2013-07-01T23:59:59.000Z

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

  11. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect (OSTI)

    Hans Gougar

    2014-05-01T23:59:59.000Z

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

  12. CERTIFICATE OF REGISTRATION Quality Management Systems

    E-Print Network [OSTI]

    Rodriguez, Carlos

    CERTIFICATE OF REGISTRATION Quality Management Systems INDIAN AGRICULTURAL STATISTICS RESEARCH Certifications Limited Certifies that the Management System of the above mentioned Company has been assessed and meets the requirements established by the following rules: ISO 9001 :2008 The Management System Includes

  13. Learning Lessons to Promote Certification and

    E-Print Network [OSTI]

    Financing and Investment in Companies Engaged in Destructive or Illegal Logging in Indonesia 14 2Learning Lessons to Promote Certification and Combat Illegal Logging in Indonesia September 2003;Learning Lessons to Promote Certification and Combat Illegal Logging in Indonesia September 2003 to June

  14. Insulin Pump Software Certification Yihai Chen1,

    E-Print Network [OSTI]

    Lawford, Mark

    Insulin Pump Software Certification Yihai Chen1, , Mark Lawford2, , Hao Wang2, , and Alan Wassyng2 for Software Certification McMaster University, Hamilton, Ontario, Canada Abstract. The insulin pump. Malfunction of the insulin pump will endanger the user's life. All countries impose some regula- tion

  15. The Graduate Certificate in Management of Technology

    E-Print Network [OSTI]

    Bieber, Michael

    The Graduate Certificate in Management of Technology Entirely Online and In Class New Jersey Institute of Technology School of Management THE EDGE IN KNOWLEDGE May 2012 Rev Use your smartphone to scan://www.njit.edu/continuing/ #12;WHY PURSUE A GRADUATE CERTIFICATE IN MANAGEMENT OF TECHNOLOGY? Developments in science

  16. Modified 2/11/2003 PATENT CERTIFICATION

    E-Print Network [OSTI]

    Pennycook, Steve

    DOE/ORO Modified 2/11/2003 PATENT CERTIFICATION Interim Certification Subcontractor Final and disclosing subject inventions as required by the patent clause of the subcontract have been followed. The completion date of this Subcontract is as follows: . The following period is covered by this patent

  17. Graduate Certificate in Construction Engineering and Management

    E-Print Network [OSTI]

    Gelfond, Michael

    , Sustainable construction methods and materials in buildings and infrastructure systems, Engineering costGraduate Certificate in Construction Engineering and Management WHY DO I NEED THIS GRADUATE The department offers a 12-hour Graduate Certificate in Construction Engineering and Management. The coursework

  18. ITEM #10(a) Proposed Academic Certificate Program

    E-Print Network [OSTI]

    O'Toole, Alice J.

    ITEM #10(a) Proposed Academic Certificate Program Title: Graduate Certificate in Product Lifecycle to an order-centric view in the 90s to product-centric view today. As product lifecycles shrink, innovation. The product lifecycle perspective becomes more important because it provides a holistic view across disparate

  19. Roundtable on Sustainable Biofuels Certification Readiness Study

    E-Print Network [OSTI]

    Roundtable on Sustainable Biofuels Certification Readiness Study: Hawai`i Biofuel Projects Prepared 12.1 Deliverable Bioenergy Analyses Prepared by Hawai`i Biofuel Foundation And NCSI Americas Inc agency thereof. #12;1 RSB Certification Readiness Study: Hawaii Biofuel Projects Prepared For Hawaii

  20. alcohol membrane reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Summary: The design and operation of a differential Polymer Electrolyte Membrane (PEM) fuel cell is described. The fuel cell design is based on coupled Stirred Tank Reactors...

  1. Global Optimization of Chemical Reactors and Kinetic Optimization

    E-Print Network [OSTI]

    ALHUSSEINI, ZAYNA ISHAQ

    2013-01-01T23:59:59.000Z

    i> plot. Chemical Engineering Research and Design. 2012; 90,Chemical Engineering Science 2002b; 57(15): Feinberg, M. Optimal reactor design

  2. ME 361E Nuclear Reactor Engineering ABET EC2000 syllabus

    E-Print Network [OSTI]

    Ben-Yakar, Adela

    ME 361E ­ Nuclear Reactor Engineering Page 1 ABET EC2000 syllabus ME 361E ­ Nuclear Reactor students should be able to: · Compare and contrast numerous nuclear reactor designs · Calculate the effects of nuclear fuel burnup · Summarize the mechanism that affect the control of a nuclear reactor Topics Covered

  3. Potential of Mirror Systems as Future Fusion Power Reactors

    SciTech Connect (OSTI)

    Kessler, Guenter; Kulcinski, Gerald L. [University of Madison (United States)

    2005-01-15T23:59:59.000Z

    Mirror based fusion reactors - as other fusion reactor concepts - have considerable environmental and safety advantages. They could make available energy resources for many 1000 years. Mirror type fusion reactors have additional technical advantages over other fusion reactor concepts. These are: simple design topology, steady state power generation, decoupling of end plugs from central power producing regions, small power units as demonstration facilities.

  4. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    SciTech Connect (OSTI)

    David E. Shropshire

    2004-04-01T23:59:59.000Z

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

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

    SciTech Connect (OSTI)

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

    2014-09-01T23:59:59.000Z

    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 needs. The experimental database will guide development of appropriate predictive methods and be available for code verification and validation (V&V) related to these systems.

  6. Quick Guide: Renewable Energy Certificates (RECs) (Fact Sheet)

    SciTech Connect (OSTI)

    Not Available

    2011-07-01T23:59:59.000Z

    Guide for Federal agencies considering renewable energy certificate (REC) purchases to fulfill Federal renewable energy requirements.

  7. Certification of a weld produced by friction stir welding

    DOE Patents [OSTI]

    Obaditch, Chris; Grant, Glenn J

    2013-10-01T23:59:59.000Z

    Methods, devices, and systems for providing certification of friction stir welds are disclosed. A sensor is used to collect information related to a friction stir weld. Data from the sensor is compared to threshold values provided by an extrinsic standard setting organizations using a certification engine. The certification engine subsequently produces a report on the certification status of the weld.

  8. Compact Proof Certificates For Linear Logic Kaustuv Chaudhuri

    E-Print Network [OSTI]

    Chaudhuri, Kaustuv

    Compact Proof Certificates For Linear Logic Kaustuv Chaudhuri INRIA, France http://kaustuv.chaudhuri

  9. A Design Study for Thermal Control of a CVD Reactor for YBCO Martha A. Gallivan, David G. Goodwin, and Richard M. Murray

    E-Print Network [OSTI]

    Gallivan, Martha A.

    chemical vapor deposition is studied using a nonlinear heat transfer model validated by preliminary Heat Transfer Model A wavelength-dependent heat transfer model of the reactor has been developed, which

  10. Self-Sustaining Thorium Boiling Water Reactors

    E-Print Network [OSTI]

    Ganda, Francesco

    A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar ...

  11. Guidelines for developing certification programs for newly generated TRU waste

    SciTech Connect (OSTI)

    Whitty, W.J.; Ostenak, C.A.; Pillay, K.K.S.; Geoffrion, R.R.

    1983-05-01T23:59:59.000Z

    These guidelines were prepared with direction from the US Department of Energy (DOE) Transuranic (TRU) Waste Management Program in support of the DOE effort to certify that newly generated TRU wastes meet the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria. The guidelines provide instructions for generic Certification Program preparation for TRU-waste generators preparing site-specific Certification Programs in response to WIPP requirements. The guidelines address all major aspects of a Certification Program that are necessary to satisfy the WIPP Waste Acceptance Criteria and their associated Compliance Requirements and Certification Quality Assurance Requirements. The details of the major element of a Certification Program, namely, the Certification Plan, are described. The Certification Plan relies on supporting data and control documentation to provide a traceable, auditable account of certification activities. Examples of specific parts of the Certification Plan illustrate the recommended degree of detail. Also, a brief description of generic waste processes related to certification activities is included.

  12. Nuclear reactor engineering

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1981-01-01T23:59:59.000Z

    Chapters are presented concerning energy from nuclear fission; nuclear reactions and radiations; diffusion and slowing-down of neutrons; principles of reactor analysis; nuclear reactor kinetics and control; energy removal; non-fuel reactor materials; the reactor fuel system; radiation protection and environmental effects; nuclear reactor shielding; nuclear reactor safety; and power reactor systems.

  13. Reactor assessments of advanced bumpy torus configurations

    SciTech Connect (OSTI)

    Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

    1983-01-01T23:59:59.000Z

    Recently, several configurational approaches and concept improvement schemes were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These configurations include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator-snakey torus). Preliminary evaluations of reactor implications of each of these configurations have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties. Results indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past.

  14. Quantifying the Reactive Uptake of OH by Organic Aerosols in a Continuous Flow Stirred Tank Reactor

    E-Print Network [OSTI]

    Che, Dung L.

    2010-01-01T23:59:59.000Z

    Continuous Flow Stirred Tank Reactor Dung L. Che, 1,2 Jaredflow stirred tank reactor. This approach is designed totimes in a flow tube reactor. This difference suggests that

  15. Radiation Shielding for Fusion Reactors

    SciTech Connect (OSTI)

    Santoro, R.T.

    1999-10-01T23:59:59.000Z

    Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel.

  16. Starfire - a commercial tokamak reactor

    SciTech Connect (OSTI)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Kokoszenski, J.; Graumann, D.

    1981-01-01T23:59:59.000Z

    The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. 10 refs.

  17. MCZ 050411 1 Stellarator Reactors

    E-Print Network [OSTI]

    designs => Motivated development of more compact designs · SPPS projected cost of electricity similar Average Major Radius (m) ASRA-6C 20 m HSR-G 18 m SPPS 14 m FFHR-J 10 m ARIES-ST Spherical Torus 3.2 m to tokamaks, but higher initial capital cost · Stellarator reactors expected to operate in true ignition

  18. CERTIFICATE

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItem NotEnergy,ARMForms About Batteries BatteriesCAES Home Home AboutCBFOCERN

  19. Graduate Certificate in Internet Business Looking for a Career in Internet Business?

    E-Print Network [OSTI]

    Pientka, Brigitte

    1 Graduate Certificate in Internet Business Looking for a Career in Internet Business? Since growth, hence a demand for experts who know how to effectively analyze and manage a business with the appropriate internet technology. Small to large-sized businesses require professionals who are able to design

  20. Automatic safety rod for reactors

    DOE Patents [OSTI]

    Germer, John H. (San Jose, CA)

    1988-01-01T23:59:59.000Z

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  1. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L. (Stanford, CA); Bachmann, Andre (Palo Alto, CA)

    1992-01-01T23:59:59.000Z

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  2. Solid State Reactor Final Report

    SciTech Connect (OSTI)

    Mays, G.T.

    2004-03-10T23:59:59.000Z

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas of research were undertaken: (1) establishing the design and safety-related basis via neutronic and reactor control assessments with the graphite foam as heat transfer medium; (2) evaluating the thermal performance of the graphite foam for heat removal, reactor stability, reactor operations, and overall core thermal characteristics; (3) characterizing the physical properties of the graphite foam under normal and irradiated conditions to determine any effects on structure, dimensional stability, thermal conductivity, and thermal expansion; and (4) developing a power conversion system design to match the reactor operating parameters.

  3. Thermomechanical analysis of fast-burst reactors

    SciTech Connect (OSTI)

    Miller, J.D.

    1994-08-01T23:59:59.000Z

    Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

  4. Application for SAS Certificate Advanced Statistics

    E-Print Network [OSTI]

    Dahl, David B.

    Application for SAS Certificate Advanced Statistics ___________________________________________________________________________________ Student Signature _________________________________________ Date I am a Statistics Major at Brigham Young of Statistics to access my academic information. Brigham Young University Department of Statistics offers

  5. Hanford Site Transuranic (TRU) Waste Certification Plan

    SciTech Connect (OSTI)

    GREAGER, T.M.

    1999-09-09T23:59:59.000Z

    The Hanford Site Transuranic Waste Certification Plan establishes the programmatic framework and criteria within which the Hanford Site ensures that contract-handled TRU wastes can be certified as compliant with the WIPP WAC and TRUPACT-II SARP.

  6. Hanford Site Transuranic (TRU) Waste Certification Plan

    SciTech Connect (OSTI)

    GREAGER, T.M.

    1999-12-14T23:59:59.000Z

    The Hanford Site Transuranic Waste Certification Plan establishes the programmatic framework and criteria with in which the Hanford Site ensures that contract-handled TRU wastes can be certified as compliant with the WIPP WAC and TRUPACT-II SARP.

  7. STATE OF CALIFORNIA CERTIFICATE OF COMPLIANCE

    E-Print Network [OSTI]

    STATE OF CALIFORNIA CERTIFICATE OF COMPLIANCE CEC-RWH-1C (Revised 08/09) CALIFORNIA ENERGY Project Name: Climate Zone: Conditioned Floor Area: Project Address: Date: General Information Building Warehouse space is Efficiency Regulations (Title 20) for walk

  8. Puerto Rico- Solar and Wind Contractor Certification

    Broader source: Energy.gov [DOE]

    In October 2008, the Energy Affairs Administration (EAA) of Puerto Rico adopted regulations for the certification of photovoltaic (PV) systems and installers in response to the passing of Act No....

  9. DOCTORATE & CERTIFICATE of the Graduate Academy

    E-Print Network [OSTI]

    Greifswald, Ernst-Moritz-Arndt-Universität

    DOCTORATE & CERTIFICATE of the Graduate Academy Own presen tations Other presen tations Key quali within the Graduate Academy Graduiertenakademie #12;Name of the Graduate Academy of the Ernst-Moritz-Arndt-University of Greifswald In addition to your own doctoral

  10. Multi-Application Small Light Water Reactor Final Report

    SciTech Connect (OSTI)

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01T23:59:59.000Z

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept. Applications such as cogeneration, water desalination or district heating were not addressed directly in the economic analyses since these depend more on local conditions, demand and economy and can not be easily generalized. Current economic performance experience and available cost data were used. The preliminary cost estimate, based on a concept that could be deployed in less than a decade, is: (1) Net Electrical Output--1050 MWe; (2) Net Station Efficiency--23%; (3) Number of Power Units--30; (4) Nominal Plant Capacity Factor--95%; (5) Total capital cost--$1241/kWe; and (6) Total busbar cost--3.4 cents/kWh. The project includes a testing program that has been conducted at Oregon State University (OSU). The test facility is a 1/3-height and 1/254.7 volume scaled design that will operate at full system pressure and temperature, and will be capable of operation at 600 kW. The design and construction of the facility have been completed. Testing is scheduled to begin in October 2002. The MASLWR conceptual design is simple, safe, and economical. It operates at NSSS parameters much lower than for a typical PWR plant, and has a much simplified power generation system. The individual reactor modules can be operated as on/off units, thereby limiting operational transients to startup and shutdown. In addition, a plant can be built in increments that match demand increases. The ''pull and replace'' concept offers automation of refueling and maintenance activities. Performing refueling in a single location improves proliferation resistance and eliminates the threat of diversion. Design certification based on testing is simplified because of the relatively low cost of a full-scale prototype facility. The overall conclusion is that while the efficiency of the power generation unit is much lower (23% versus 30%), the reduction in capital cost due to simplification of design more than makes up for the increased cost of nuclear fuel. The design concept complies with the safety requirements and criteria. It also satisfies the goals for modularity, standard plant design, certification before construction, c

  11. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    SciTech Connect (OSTI)

    Bernard A. Toseland

    2002-09-30T23:59:59.000Z

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors.

  12. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    SciTech Connect (OSTI)

    Bernard A. Toseland, Ph.D.

    2000-06-01T23:59:59.000Z

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column 0reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors.

  13. Advanced reactor safety research, quarterly report, October-December 1980

    SciTech Connect (OSTI)

    Not Available

    1982-01-01T23:59:59.000Z

    Information is presented concerning advanced reactor core phenomenology; light water reactor severe core damage phenomenology; core debris behavior; containment analysis; elevated temperature design assessment; LMFBR accident delineation; and test and facility technology.

  14. Engineering Development of Slurry Bubble Column Reactor (SBCR) Technology

    SciTech Connect (OSTI)

    Toseland, B.A.

    1998-10-29T23:59:59.000Z

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors.

  15. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    SciTech Connect (OSTI)

    Bernard A. Toseland, Ph.D.

    1999-03-01T23:59:59.000Z

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors. The past three months of research have been focused on two major areas of bubble column hydrodynamics: (1) pressure and temperature effects on gas holdup and (2) region transition using a sparger as a gas distributor.

  16. atucha-2 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  17. atucha-1 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  18. acpr reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  19. atlantic-1 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 6 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  20. arbus reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  1. akm reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 9 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  2. argonaut reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 14 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  3. argonaut lemont reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 15 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  4. arkansas-2 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  5. apra reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 7 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  6. apsara reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 6 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  7. asco-1 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  8. atlantic-2 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  9. akr-1 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 6 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  10. akm muehleberg reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 9 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  11. arsi reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 12 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  12. aipfr reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  13. asco-2 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  14. argonauta rien-1 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 6 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  15. ahfr reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  16. almaraz-2 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  17. aarr reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  18. almaraz-1 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  19. arkansas-1 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 6 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  20. akw1 rheinsberg reactor: Topics by E-print Network

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    the flux calculation and recent progresses. Jun Cao 2012-03-08 6 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  1. aguirre-1 reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 11 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  2. arbi reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 7 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  3. agesta reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  4. acrr reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  5. affri reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  6. afsr reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  7. atsr reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 7 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  8. alrr reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    the flux calculation and recent progresses. Jun Cao 2012-03-08 6 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  9. Development of a system model for advanced small modular reactors.

    SciTech Connect (OSTI)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01T23:59:59.000Z

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandia's concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  10. Cross section generation strategy for high conversion light water reactors

    E-Print Network [OSTI]

    Herman, Bryan R. (Bryan Robert)

    2011-01-01T23:59:59.000Z

    High conversion water reactors (HCWR), such as the Resource-renewable Boiling Water Reactor (RBWR), are being designed with axial heterogeneity of alternating fissile and blanket zones to achieve a conversion ratio of ...

  11. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect (OSTI)

    R. Wigeland; J. Cahalan

    2009-09-01T23:59:59.000Z

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

  12. Radioisotope thermoelectric generator licensed hardware package and certification tests

    SciTech Connect (OSTI)

    Goldmann, L.H.; Averette, H.S. [Westinghouse Hanford Company, P.O. Box 1970, M/S R3-86 or N1-32, Richland, Washington 99352 (United States)

    1995-01-20T23:59:59.000Z

    This paper presents the Licensed Hardware package and the Certification Test portions of the Radioisitope Themoelectric Generator Transportation System. This package has been designed to meet those portions of the {ital Code} {ital of} {ital Federal} {ital Regulations} (10 CFR 71) relating to ``Type B`` shipments of radioactive materials. The licensed hardware is now in the U. S. Department of Energy licensing process that certifies the packaging`s integrity under accident conditions. The detailed information for the anticipated license is presented in the safety analysis report for packaging, which is now in process and undergoing necessary reviews. As part of the licensing process, a full-size Certification Test Article unit, which has modifications slightly different than the Licensed Hardware or production shipping units, is used for testing. Dimensional checks of the Certification Test Article were made at the manufacturing facility. Leak testing and drop testing were done at the 300 Area of the U.S. Department of Energy`s Hanford Site near Richland, Washington. The hardware includes independent double containments to prevent the environmental spread of {sup 238}Pu, impact limiting devices to protect portions of the package from impacts, and thermal insulation to protect the seal areas from excess heat during accident conditions. The package also features electronic feed-throughs to monitor the Radioisotope Thermoelectric Generator`s temperature inside the containment during the shipment cycle. This package is designed to safely dissipate the typical 4,500 thermal watts produced in the largest Radioisotope Thermoelectric Generators. The package also contains provisions to ensure leak tightness when radioactive materials, such as a Radioisotope Thermoelectric Generator for the Cassini Mission, planned for 1997 by the National Aeronautics and Space Administration, are being prepared for shipment. (Abstract Truncated)

  13. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

    1983-01-01T23:59:59.000Z

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  14. Laser Inertial Fusion-based Energy: Neutronic Design Aspects of a Hybrid Fusion-Fission Nuclear Energy System

    E-Print Network [OSTI]

    Kramer, Kevin James

    2010-01-01T23:59:59.000Z

    IV example sodium fast reactor. assessing design variations.generate fuel for fast nuclear reactors, although Basov and

  15. Certification of the Cessna 152 on 100% ethanol

    SciTech Connect (OSTI)

    Shauck, M.E.; Zanin, M.G.

    1997-12-31T23:59:59.000Z

    In June 1996, the Renewable Aviation Fuels Development Center (RAFDC) at Baylor University in Waco, Texas, received a Supplemental Type Certificate (STC) for the use of 100% ethanol as a fuel for the Cessna 152, the most popular training aircraft in the world. This is the first certification granted by the Federal Aviation Administration (FAA) for a non-petroleum fuel. Certification of an aircraft on a new fuel requires a certification of the engine followed by a certification of the airframe/engine combination. This paper will describe the FAA airframe certification procedure, the tests required and their outcome using ethanol as an aviation fuel in a Cessna 152.

  16. Reduced Order Based Compensator Control of Thin Film Growth in a CVD Reactor

    E-Print Network [OSTI]

    Reduced Order Based Compensator Control of Thin Film Growth in a CVD Reactor H.T. Banks and H chemical vapor deposition (CVD) reactors. An in­ tegral component of this research program is the design of the reactor so that control and sensing are a basic component of the optimal design e#orts for the reactor. We

  17. Department of Earth and Mineral Engineering Spring 2011 Oxidative Coupling of Methane Reactor

    E-Print Network [OSTI]

    Demirel, Melik C.

    Reactor Overview The sponsor has tasked our team with the design, construction and testing of an experimental reactor designed to couple methane to ethane and dehydrogenate ethane to ethylene. The reactor and build the reactor and perform methane conversion testing to provide proof of concept for the OCM

  18. Reduced Order Based Compensator Control of Thin Film Growth in a CVD Reactor

    E-Print Network [OSTI]

    Reduced Order Based Compensator Control of Thin Film Growth in a CVD Reactor H.T. Banks and H chemical vapor deposition (CVD) reactors. An in- tegral component of this research program is the design of the reactor so that control and sensing are a basic component of the optimal design efforts for the reactor

  19. Nuclear reactor safety heat transfer

    SciTech Connect (OSTI)

    Jones, O.C.

    1982-07-01T23:59:59.000Z

    Reviewed is a book which has 5 parts: Overview, Fundamental Concepts, Design Basis Accident-Light Water Reactors (LWRs), Design Basis Accident-Liquid-Metal Fast Breeder Reactors (LMFBRs), and Special Topics. It combines a historical overview, textbook material, handbook information, and the editor's personal philosophy on safety of nuclear power plants. Topics include thermal-hydraulic considerations; transient response of LWRs and LMFBRs following initiating events; various accident scenarios; single- and two-phase flow; single- and two-phase heat transfer; nuclear systems safety modeling; startup and shutdown; transient response during normal and upset conditions; vapor explosions, natural convection cooling; blockages in LMFBR subassemblies; sodium boiling; and Three Mile Island.

  20. Modular Inspection System for a Complete IN-Service Examination of Nuclear Reactor Pressure Vessel, Including Beltline Region

    SciTech Connect (OSTI)

    David H. Bothell

    2000-04-30T23:59:59.000Z

    Final Report for a DOE Phase II Contract Describing the design and fabrication of a reactor inspection modular rover prototype for reactor vessel inspection.