National Library of Energy BETA

Sample records for modular reactor designs

  1. Modular stellarator reactor conceptual design study

    SciTech Connect (OSTI)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.

    1983-01-01

    A conceptual design study of the Modular Stellarator Reactor is summarized. The physics basis of the approach is elucidated with emphasis on magnetics performance optimization. Key engineering features of the fusion power core are described. Comparisons with an analogous continuous-helical-coil (torsatron) system are made as the basis of a technical and economic assessment.

  2. Generic small modular reactor plant design.

    SciTech Connect (OSTI)

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

    2012-12-01

    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.

  3. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  4. Modularity in design of the MIT Pebble Bed Reactor

    E-Print Network [OSTI]

    Berte, Marc Vincent, 1977-

    2004-01-01

    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 ...

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

    E-Print Network [OSTI]

    Design, Analysis and Optimization of the Power Conversion System for the Modular Pebble Bed Reactor August 31, 2003 Certified by Ronald G. Ballinger Professor of Nuclear Engineering and Materials Science;2 #12;Design, Analysis, and Optimization of the Power Conversion System for the Modular Pebble Bed

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

    E-Print Network [OSTI]

    Everson, Matthew S

    2009-01-01

    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 ...

  7. 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.

    and accepted September 10, 2002) A metal fueled modular island core sodium cooled fast breeder reactor concept Design of a Modular Island Core Fast Breeder Reactor "RAPID-M" Mitsuru KAMBE Central Research Institute type reactors, performance, sodium cooled reactor, modular island core, inherent safety, in- tegrated

  8. Advanced Modularity Design for The MIT Pebble Bed Reactor

    E-Print Network [OSTI]

    be economically built and operated. One of the major impediments to new nuclear construction is the capital costs small size the capital cost per kilowatt is likely to be large if traditional construction approaches Reactor 1. Introduction The capital cost of new nuclear plants is the most significant reason for the lack

  9. 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

    2010-01-01

    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.

  10. Modular Stellarator Fusion Reactor concept

    SciTech Connect (OSTI)

    Miller, R.L.; Krakowski, R.A.

    1981-08-01

    A preliminary conceptual study is made of the Modular Stellarator Reactor (MSR). A steady-state ignited, DT-fueled, magnetic fusion reactor is proposed for use as a central electric-power station. The MSR concept combines the physics of the classic stellarator confinement topology with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. The physics basis of the design point is described together with supporting magnetics, coil-force, and stress computations. The approach and results presented herein will be modified in the course of ongoing work to form a firmer basis for a detailed conceptual design of the MSR.

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

    E-Print Network [OSTI]

    Wang, Chunyun, 1968-

    2003-01-01

    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 ...

  12. 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

    2006-06-01

    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.

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

    SciTech Connect (OSTI)

    Memmott, M. J.; Stansbury, C.; Taylor, C.

    2012-07-01

    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)

  14. 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

    2003-06-16

    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

  15. 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

    2013-10-01

    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.

  16. Modularity Approach Modular Pebble Bed Reactor (MPBR)

    E-Print Network [OSTI]

    NED MPBR 1150 MW Combined Heat and Power Station Turbine Hall Boundary Admin Training Control Bldg. · No Reprocessing · High Burnup >90,000 Mwd/MT · Direct Disposal of HLW · Process Heat Applications - Hydrogen · On--line Refueling #12;4/23/03 MIT NED MPBR Reference Plant Modular Pebble Bed Reactor Thermal Power

  17. Small Modular Reactors: Institutional Assessment

    SciTech Connect (OSTI)

    Joseph Perkowski, Ph.D.

    2012-06-01

    ? Objectives include, among others, a description of the basic development status of “small modular reactors” (SMRs) focused primarily on domestic activity; investigation of the domestic market appeal of modular reactors from the viewpoints of both key energy sector customers and also key stakeholders in the financial community; and consideration of how to proceed further with a pro-active "core group" of stakeholders substantially interested in modular nuclear deployment in order to provide the basis to expedite design/construction activity and regulatory approval. ? Information gathering was via available resources, both published and personal communications with key individual stakeholders; published information is limited to that already in public domain (no confidentiality); viewpoints from interviews are incorporated within. Discussions at both government-hosted and private-hosted SMR meetings are reflected herein. INL itself maintains a neutral view on all issues described. Note: as per prior discussion between INL and CAP, individual and highly knowledgeable senior-level stakeholders provided the bulk of insights herein, and the results of those interviews are the main source of the observations of this report. ? Attachment A is the list of individual stakeholders consulted to date, including some who provided significant earlier assessments of SMR institutional feasibility. ? Attachments B, C, and D are included to provide substantial context on the international status of SMR development; they are not intended to be comprehensive and are individualized due to the separate nature of the source materials. Attachment E is a summary of the DOE requirements for winning teams regarding the current SMR solicitation. Attachment F deserves separate consideration due to the relative maturity of the SMART SMR program underway in Korea. Attachment G provides illustrative SMR design features and is intended for background. Attachment H is included for overview purposes and is a sampling of advanced SMR concepts, which will be considered as part of the current DOE SMR program but whose estimated deployment time is beyond CAP’s current investment time horizon. Attachment I is the public DOE statement describing the present approach of their SMR Program.

  18. Modular stellarator reactor: a fusion power plant

    SciTech Connect (OSTI)

    Miller, R.L.; Bathke, C.G.; Krakowski, R.A.; Heck, F.M.; Green, L.; Karbowski, J.S.; Murphy, J.H.; Tupper, R.B.; DeLuca, R.A.; Moazed, A.

    1983-07-01

    A comparative analysis of the modular stellarator and the torsatron concepts is made based upon a steady-state ignited, DT-fueled, reactor embodiment of each concept for use as a central electric-power station. Parametric tradeoff calculations lead to the selection of four design points for an approx. 4-GWt plant based upon Alcator transport scaling in l = 2 systems of moderate aspect ratio. The four design points represent high-aspect ratio. The four design points represent high-(0.08) and low-(0.04) beta versions of the modular stellarator and torsatron concepts. The physics basis of each design point is described together with supporting engineering and economic analyses. The primary intent of this study is the elucidation of key physics and engineering tradeoffs, constraints, and uncertainties with respect to the ultimate power reactor embodiment.

  19. Reactor Engineering Design | netl.doe.gov

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

    Reactor Engineering Design The Reactor Engineering Design Key Technology will focus on control of chemical reactions with unprecedented precision in increasingly modular and...

  20. Small Modular Reactors (468th Brookhaven Lecture)

    SciTech Connect (OSTI)

    Bari, Robert

    2011-04-20

    With good reason, much more media attention has focused on nuclear power plants than solar farms, wind farms, or hydroelectric plants during the past month and a half. But as nations around the world demand more energy to power everything from cell phone batteries to drinking water pumps to foundries, nuclear plants are the only non-greenhouse-gas producing option that can be built to operate almost anywhere, and can continue to generate power during droughts, after the sun sets, and when winds die down. To supply this demand for power, designers around the world are competing to develop more affordable nuclear reactors of the future: small modular reactors. Brookhaven Lab is working with DOE to ensure that these reactors are designed to be safe for workers, members of surrounding communities, and the environment and to ensure that the radioactive materials and technology will only be used for peaceful purposes, not weapons. In his talk, Bari will discuss the advantages and challenges of small modular reactors and what drives both international and domestic interest in them. He will also explain how Brookhaven Lab and DOE are working to address the challenges and provide a framework for small modular reactors to be commercialized.

  1. Energy Department Announces Small Modular Reactor Technology...

    Energy Savers [EERE]

    today three public-private partnerships to develop deployment plans for small modular nuclear reactor (SMR) technologies at SRS facilities, near Aiken, South Carolina. As part...

  2. Development of a system model for advanced small modular reactors.

    SciTech Connect (OSTI)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    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.

  3. Proliferation resistance of small modular reactors fuels

    SciTech Connect (OSTI)

    Polidoro, F.; Parozzi, F.; Fassnacht, F.; Kuett, M.; Englert, M.

    2013-07-01

    In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

  4. MODULAR PEBBLE BED REACTOR PROJECT UNIVERSITY RESEARCH CONSORTIUM

    E-Print Network [OSTI]

    includes the development of a fission gas release model, particle temperature distributions, internal conceptual design for a gas-cooled, modular, pebble bed reactor. Key technology areas associated particle pressure, migration of fission products, and chemical attack of fuel particle layers. · A balance

  5. Computational Analysis of Fluid Flow in Pebble Bed Modular Reactor 

    E-Print Network [OSTI]

    Gandhir, Akshay

    2012-10-19

    High Temperature Gas-cooled Reactor (HTGR) is a Generation IV reactor under consideration by Department of Energy and in the nuclear industry. There are two categories of HTGRs, namely, Pebble Bed Modular Reactor (PBMR) and Prismatic reactor. Pebble...

  6. Small modular reactors (SMRs) such...

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

    (ARO), using soluble boron in the coolant for reactivity control. Conversely, boiling water reactors (BWRs) typically maneuver their control blades as often as every 2 GWdmtU...

  7. Human Reliability Considerations for Small Modular Reactors

    SciTech Connect (OSTI)

    OHara J. M.; Higgins, H.; DAgostino, A.; Erasmia, L.

    2012-01-27

    Small modular reactors (SMRs) are a promising approach to meeting future energy needs. Although the electrical output of an individual SMR is relatively small compared to that of typical commercial nuclear plants, they can be grouped to produce as much energy as a utility demands. Furthermore, SMRs can be used for other purposes, such as producing hydrogen and generating process heat. The design characteristics of many SMRs differ from those of current conventional plants and may require a distinct concept of operations. The U.S. Nuclear Regulatory Commission (NRC) conducted research to examine the human factors engineering and the operational aspects of SMRs. The research identified thirty potential human-performance issues that should be considered in the NRC's reviews of SMR designs and in future research activities. The purpose of this report is to illustrate how the issues can support SMR probabilistic risk analyses and their review by identifying potential human failure events for a subset of the issues. As part of addressing the human contribution to plant risk, human reliability analysis practitioners identify and quantify the human failure events that can negatively impact normal or emergency plant operations. The results illustrated here can be generalized to identify additional human failure events for the issues discussed and can be applied to those issues not discussed in this report.

  8. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    SciTech Connect (OSTI)

    E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

    2010-09-20

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using factory prefabricated structural modules, for application to external event shell and base isolated structures.

  9. Development and Optimization of Modular Hybrid Plasma Reactor...

    Office of Scientific and Technical Information (OSTI)

    Optimization of Modular Hybrid Plasma Reactor N A 36 MATERIALS SCIENCE INL developed a bench-scale, modular hybrid plasma system for gas-phase nanomaterials synthesis. The system...

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

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01

    safety design criteria separate effects test steam generators small modular reactor San Onofre Nuclear

  11. Modular hybrid plasma reactor and related systems and methods...

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

    Patent Search Success Stories News Events Find More Like This Return to Search Modular hybrid plasma reactor and related systems and methods United States Patent Patent Number:...

  12. Modular hybrid plasma reactor and related systems and methods...

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

    (27) Visual Patent Search Success Stories News Events Return to Search Modular hybrid plasma reactor and related systems and methods United States Patent Application ***...

  13. Small modular reactors (SMRs) such as the

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home RoomPreservation ofAlbuquerque|Sensitive Species3performed Steven D. PainSmall modular reactors

  14. Advanced Small Modular Reactor Economics Model Development

    SciTech Connect (OSTI)

    Harrison, Thomas J.

    2014-10-01

    The US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (SMR) research and development activities focus on four key areas: Developing assessment methods for evaluating advanced SMR technologies and characteristics; and Developing and testing of materials, fuels and fabrication techniques; and Resolving key regulatory issues identified by US Nuclear Regulatory Commission and industry; and Developing advanced instrumentation and controls and human-machine interfaces. This report focuses on development of assessment methods to evaluate advanced SMR technologies and characteristics. Specifically, this report describes the expansion and application of the economic modeling effort at Oak Ridge National Laboratory. Analysis of the current modeling methods shows that one of the primary concerns for the modeling effort is the handling of uncertainty in cost estimates. Monte Carlo–based methods are commonly used to handle uncertainty, especially when implemented by a stand-alone script within a program such as Python or MATLAB. However, a script-based model requires each potential user to have access to a compiler and an executable capable of handling the script. Making the model accessible to multiple independent analysts is best accomplished by implementing the model in a common computing tool such as Microsoft Excel. Excel is readily available and accessible to most system analysts, but it is not designed for straightforward implementation of a Monte Carlo–based method. Using a Monte Carlo algorithm requires in-spreadsheet scripting and statistical analyses or the use of add-ons such as Crystal Ball. An alternative method uses propagation of error calculations in the existing Excel-based system to estimate system cost uncertainty. This method has the advantage of using Microsoft Excel as is, but it requires the use of simplifying assumptions. These assumptions do not necessarily bring into question the analytical results. In fact, the analysis shows that the propagation of error method introduces essentially negligible error, especially when compared to the uncertainty associated with some of the estimates themselves. The results of these uncertainty analyses generally quantify and identify the sources of uncertainty in the overall cost estimation. The obvious generalization—that capital cost uncertainty is the main driver—can be shown to be an accurate generalization for the current state of reactor cost analysis. However, the detailed analysis on a component-by-component basis helps to demonstrate which components would benefit most from research and development to decrease the uncertainty, as well as which components would benefit from research and development to decrease the absolute cost.

  15. Modular Pebble Bed Reactor High Temperature Gas Reactor

    E-Print Network [OSTI]

    For 1150 MW Combined Heat and Power Station Oil Refinery Hydrogen Production Desalinization Plant VHTR · Modularity Design · Intermediate Heat Exchanger Design · Core Power Distribution Monitoring · Pebble Flow Burnup >90,000 Mwd/MT · Direct Disposal of HLW · Process Heat Applications - Hydrogen, water #12;Turbine

  16. An Overview of the Safety Case for Small Modular Reactors

    SciTech Connect (OSTI)

    Ingersoll, Daniel T [ORNL] [ORNL

    2011-01-01

    Several small modular reactor (SMR) designs emerged in the late 1970s and early 1980s in response to lessons learned from the many technical and operational challenges of the large Generation II light-water reactors. After the accident at the Three Mile Island plant in 1979, an ensuing reactor redesign effort spawned the term inherently safe designs, which later evolved into passively safe terminology. Several new designs were engineered to be deliberately small in order to fully exploit the benefits of passive safety. Today, new SMR designs are emerging with a similar philosophy of offering highly robust and resilient designs with increased safety margins. Additionally, because these contemporary designs are being developed subsequent to the September 11, 2001, terrorist attack, they incorporate a number of intrinsic design features to further strengthen their safety and security. Several SMR designs are being developed in the United States spanning the full spectrum of reactor technologies, including water-, gas-, and liquid-metal-cooled ones. Despite a number of design differences, most of these designs share a common set of design principles to enhance plant safety and robustness, such as eliminating plant design vulnerabilities where possible, reducing accident probabilities, and mitigating accident consequences. An important consequence of the added resilience provided by these design approaches is that the individual reactor units and the entire plant should be able to survive a broader range of extreme conditions. This will enable them to not only ensure the safety of the general public but also help protect the investment of the owner and continued availability of the power-generating asset. Examples of typical SMR design features and their implications for improved plant safety are given for specific SMR designs being developed in the United States.

  17. ASSESSMENT OF SMALL AND MODULAR REACTOR NUCLEAR FUEL COST 

    E-Print Network [OSTI]

    Pannier, Christopher 1992-

    2012-05-03

    INCAS INtegrated model for the Competitiveness Analysis of Small modular reactors LWR Light Water Reactor NEI Nuclear Energy Institute PWR Pressurized Water Reactor PHWR Pressurized Heavy Water Reactor SEMER Système d’Évaluation et de Modélisation... ...................................................... 27 8 LWR Fuel Cost ..................................................................................................... 28 9 SMR Fuel Cost ..................................................................................................... 29...

  18. Human Reliability Analysis for Small Modular Reactors

    SciTech Connect (OSTI)

    Ronald L. Boring; David I. Gertman

    2012-06-01

    Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.

  19. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    SciTech Connect (OSTI)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next Generation Safeguards Initiative (NGSI).

  20. Prognostics Health Management for Advanced Small Modular Reactor Passive Components

    SciTech Connect (OSTI)

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

    2013-10-18

    In the United States, sustainable nuclear power to promote energy security is a key national energy priority. Advanced small modular reactors (AdvSMR), which are based on modularization of advanced reactor concepts using non-light-water reactor (LWR) coolants such as liquid metal, helium, or liquid salt may provide a longer-term alternative to more conventional LWR-based concepts. The economics of AdvSMRs will be impacted by the reduced economy-of-scale savings when compared to traditional LWRs and the controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance costs. Therefore, achieving the full benefits of AdvSMR deployment requires a new paradigm for plant design and management. In this context, prognostic health management of passive components in AdvSMRs can play a key role in enabling the economic deployment of AdvSMRs. In this paper, the background of AdvSMRs is discussed from which requirements for PHM systems are derived. The particle filter technique is proposed as a prognostics framework for AdvSMR passive components and the suitability of the particle filter technique is illustrated by using it to forecast thermal creep degradation using a physics-of-failure model and based on a combination of types of measurements conceived for passive AdvSMR components.

  1. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect (OSTI)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under 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.

  2. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect (OSTI)

    Gougar, Hans D.

    2014-10-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under 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.

  3. Advanced Small Modular Reactor Economics Status Report

    SciTech Connect (OSTI)

    Harrison, Thomas J.

    2014-10-01

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

  4. Site Suitability and Hazard Assessment Guide for Small Modular Reactors

    SciTech Connect (OSTI)

    Wayne Moe

    2013-10-01

    Commercial nuclear reactor projects in the U.S. have traditionally employed large light water reactors (LWR) to generate regional supplies of electricity. Although large LWRs have consistently dominated commercial nuclear markets both domestically and abroad, the concept of small modular reactors (SMRs) capable of producing between 30 MW(t) and 900 MW(t) to generating steam for electricity is not new. Nor is the idea of locating small nuclear reactors in close proximity to and in physical connection with industrial processes to provide a long-term source of thermal energy. Growing problems associated continued use of fossil fuels and enhancements in efficiency and safety because of recent advancements in reactor technology suggest that the likelihood of near-term SMR technology(s) deployment at multiple locations within the United States is growing. Many different types of SMR technology are viable for siting in the domestic commercial energy market. However, the potential application of a particular proprietary SMR design will vary according to the target heat end-use application and the site upon which it is proposed to be located. Reactor heat applications most commonly referenced in connection with the SMR market include electric power production, district heating, desalinization, and the supply of thermal energy to various processes that require high temperature over long time periods, or a combination thereof. Indeed, the modular construction, reliability and long operational life purported to be associated with some SMR concepts now being discussed may offer flexibility and benefits no other technology can offer. Effective siting is one of the many early challenges that face a proposed SMR installation project. Site-specific factors dealing with support to facility construction and operation, risks to the plant and the surrounding area, and the consequences subsequent to those risks must be fully identified, analyzed, and possibly mitigated before a license will be granted to construct and operate a nuclear facility. Examples of significant site-related concerns include area geotechnical and geological hazard properties, local climatology and meteorology, water resource availability, the vulnerability of surrounding populations and the environmental to adverse effects in the unlikely event of radionuclide release, the socioeconomic impacts of SMR plant installation and the effects it has on aesthetics, proximity to energy use customers, the topography and area infrastructure that affect plant constructability and security, and concerns related to the transport, installation, operation and decommissioning of major plant components.

  5. Energy Department Announces New Investment in U.S. Small Modular...

    Energy Savers [EERE]

    New Investment in U.S. Small Modular Reactor Design and Commercialization Energy Department Announces New Investment in U.S. Small Modular Reactor Design and Commercialization...

  6. Advanced Control and Protection system Design Methods for Modular HTGRs

    SciTech Connect (OSTI)

    Ball, Sydney J; Wilson Jr, Thomas L; Wood, Richard Thomas

    2012-06-01

    The project supported the Nuclear Regulatory Commission (NRC) in identifying and evaluating the regulatory implications concerning the control and protection systems proposed for use in the Department of Energy's (DOE) Next-Generation Nuclear Plant (NGNP). The NGNP, using modular high-temperature gas-cooled reactor (HTGR) technology, is to provide commercial industries with electricity and high-temperature process heat for industrial processes such as hydrogen production. Process heat temperatures range from 700 to 950 C, and for the upper range of these operation temperatures, the modular HTGR is sometimes referred to as the Very High Temperature Reactor or VHTR. Initial NGNP designs are for operation in the lower temperature range. The defining safety characteristic of the modular HTGR is that its primary defense against serious accidents is to be achieved through its inherent properties of the fuel and core. Because of its strong negative temperature coefficient of reactivity and the capability of the fuel to withstand high temperatures, fast-acting active safety systems or prompt operator actions should not be required to prevent significant fuel failure and fission product release. The plant is designed such that its inherent features should provide adequate protection despite operational errors or equipment failure. Figure 1 shows an example modular HTGR layout (prismatic core version), where its inlet coolant enters the reactor vessel at the bottom, traversing up the sides to the top plenum, down-flow through an annular core, and exiting from the lower plenum (hot duct). This research provided NRC staff with (a) insights and knowledge about the control and protection systems for the NGNP and VHTR, (b) information on the technologies/approaches under consideration for use in the reactor and process heat applications, (c) guidelines for the design of highly integrated control rooms, (d) consideration for modeling of control and protection system designs for VHTR, and (e) input for developing the bases for possible new regulatory guidance to assist in the review of an NGNP license application. This NRC project also evaluated reactor and process heat application plant simulation models employed in the protection and control system designs for various plant operational modes and accidents, including providing information about the models themselves, and the appropriateness of the application of the models for control and protection system studies. A companion project for the NRC focused on the potential for new instrumentation that would be unique to modular HTGRs, as compared to light-water reactors (LWRs), due to both the higher temperature ranges and the inherent safety features.

  7. Hybrid energy systems (HESs) using small modular reactors (SMRs)

    SciTech Connect (OSTI)

    S. Bragg-Sitton

    2014-10-01

    Large-scale nuclear reactors are traditionally operated for a singular purpose: steady-state production of dispatchable baseload electricity that is distributed broadly on the electric grid. While this implementation is key to a sustainable, reliable energy grid, small modular reactors (SMRs) offer new opportunities for increased use of clean nuclear energy for both electric and thermal ap plications in more locations – while still accommodating the desire to support renewable production sources.

  8. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    SciTech Connect (OSTI)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  9. Health Monitoring to Support Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-08-01

    Advanced small modular reactors (aSMRs) are based on advanced reactor concepts, some of which were promoted by the Generation IV International Forum, and are being considered for diverse missions including desalination of water, production of hydrogen, etc. While the existing fleet of commercial nuclear reactors provides baseload electricity, it is conceivable that aSMRs could be implemented for both baseload and load following applications. The effect of diverse operating missions and unit modularity on plant operations and maintenance (O&M) is not fully understood and limiting these costs will be essential to successful deployment of aSMRs. Integrated health monitoring concepts are proposed to support the safe and affordable operation of aSMRs over their lifetime by enabling management of significant in-vessel and in-containment active and passive components.

  10. Modular Pebble Bed Reactor March 22, 2000

    E-Print Network [OSTI]

    ;"Naturally" Safe Fuel · Shut Off All Cooling · Withdraw All Control Rods · No Emergency Cooling · No Operator · No melt down · No significant radiation release in accident · Demonstrate with actual test of reactor #12

  11. Prismatic modular reactor analysis with melcor 

    E-Print Network [OSTI]

    Zhen, Ni

    2009-05-15

    core reactor, was simulated with the complete model. MELCOR has been demonstrated to have the ability of modeling a prismatic core VHTR. The calculated outlet temperature and mass flow rate under normal operation correspond well to references. However...

  12. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Demonstration

    SciTech Connect (OSTI)

    Curtis Smith; Steven Prescott; Tony Koonce

    2014-04-01

    A key area of the Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) strategy is the development of methodologies and tools that will be used to predict the safety, security, safeguards, performance, and deployment viability of SMRs. The goal of the SMR PRA activity will be to develop quantitative methods and tools and the associated analysis framework for assessing a variety of risks. Development and implementation of SMR-focused safety assessment methods may require new analytic methods or adaptation of traditional methods to the advanced design and operational features of SMRs. We will need to move beyond the current limitations such as static, logic-based models in order to provide more integrated, scenario-based models based upon predictive modeling which are tied to causal factors. The development of SMR-specific safety models for margin determination will provide a safety case that describes potential accidents, design options (including postulated controls), and supports licensing activities by providing a technical basis for the safety envelope. This report documents the progress that was made to implement the PRA framework, specifically by way of demonstration of an advanced 3D approach to representing, quantifying and understanding flooding risks to a nuclear power plant.

  13. Modularity of the MIT Pebble Bed Reactor for use by the commercial power industry

    E-Print Network [OSTI]

    Hanlon-Hyssong, Jaime E

    2008-01-01

    The Modular Pebble Bed Reactor is a small high temperature helium cooled reactor that is being considered for both electric power and hydrogen production. Pebble bed reactors are being developed in South Africa, China and ...

  14. On Enhancing Risk Monitors for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Coles, Garill A.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-08-01

    Advanced small modular reactors (AdvSMRs) can contribute to safe, sustainable, and carbon-neutral energy production. However, the economics of AdvSMRs suffer from the loss of economy-of-scale for both construction and operation. The controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance (O&M) costs. These expenses could potentially be managed through optimized scheduling of O&M activities for components, reactor modules, power blocks, and the full plant. Accurate, real-time risk assessment with integrated health monitoring of key active components can support scheduling of both online and offline inspection and maintenance activities.

  15. Deep-Burn Modular Helium Reactor Fuel Development Plan

    SciTech Connect (OSTI)

    McEachern, D

    2002-12-02

    This document contains the workscope, schedule and cost for the technology development tasks needed to satisfy the fuel and fission product transport Design Data Needs (DDNs) for the Gas Turbine-Modular Helium Reactor (GT-MHR), operating in its role of transmuting transuranic (TRU) nuclides in spent fuel discharged from commercial light-water reactors (LWRs). In its application for transmutation, the GT-MHR is referred to as the Deep-Burn MHR (DB-MHR). This Fuel Development Plan (FDP) describes part of the overall program being undertaken by the U.S. Department of Energy (DOE), utilities, and industry to evaluate the use of the GT-MHR to transmute transuranic nuclides from spent nuclear fuel. The Fuel Development Plan (FDP) includes the work on fuel necessary to support the design and licensing of the DB-MHR. The FDP is organized into ten sections. Section 1 provides a summary of the most important features of the plan, including cost and schedule information. Section 2 describes the DB-MHR concept, the features of its fuel and the plan to develop coated particle fuel for transmutation. Section 3 describes the knowledge base for fabrication of coated particles, the experience with irradiation performance of coated particle fuels, the database for fission product transport in HTGR cores, and describes test data and calculations for the performance of coated particle fuel while in a repository. Section 4 presents the fuel performance requirements in terms of as-manufactured quality and performance of the fuel coatings under irradiation and accident conditions. These requirements are provisional because the design of the DB-MHR is in an early stage. However, the requirements are presented in this preliminary form to guide the initial work on the fuel development. Section 4 also presents limits on the irradiation conditions to which the coated particle fuel can be subjected for the core design. These limits are based on past irradiation experience. Section 5 describes the Design Data Needs to: (1) fabricate the coated particle fuel, (2) predict its performance in the reactor core, (3) predict the radionuclide release rates from the reactor core, and (4) predict the performance of spent fuel in a geological repository. The heart of this fuel development plan is Section 6, which describes the development activities proposed to satisfy the DDNs presented in Section 5. The development scope is divided into Fuel Process Development, Fuel Materials Development, Fission Product Transport, and Spent Fuel Disposal. Section 7 describes the facilities to be used. Generally, this program will utilize existing facilities. While some facilities will need to be modified, there is no requirement for major new facilities. Section 8 states the Quality Assurance requirements that will be applied to the development activities. Section 9 presents detailed costs organized by WBS and spread over time. Section 10 presents a list of the types of deliverables that will be prepared in each of the WBS elements. Four Appendices contain supplementary information on: (a) design data needs, (b) the interface with the separations plant, (c) the detailed development schedule, and (d) the detailed cost estimate.

  16. Modular design of biological systems

    E-Print Network [OSTI]

    Norville, Julie Erin, 1980-

    2012-01-01

    The focus of my research is the development of technology for building compound biological systems from simpler pieces. I designed BioScaffold parts, a family of variable regions that can be inserted into a DNA sequence ...

  17. Studies on the closed-loop digital control of multi-modular reactors

    SciTech Connect (OSTI)

    Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (United States). Nuclear Reactor Lab.); Henry, A.F.; Lanning, D.D.; Meyer, J.E. (Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering)

    1992-11-01

    This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

  18. Studies on the closed-loop digital control of multi-modular reactors. Final report

    SciTech Connect (OSTI)

    Bernard, J.A. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Nuclear Reactor Lab.; Henry, A.F.; Lanning, D.D.; Meyer, J.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering

    1992-11-01

    This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

  19. A TAXONOMY OF MODULAR GRIME IN DESIGN PATTERNS Travis Steven Schanz

    E-Print Network [OSTI]

    Dyer, Bill

    A TAXONOMY OF MODULAR GRIME IN DESIGN PATTERNS by Travis Steven Schanz A thesis submitted .......................................................................................21 3. MODULAR GRIME TAXONOMY .........................

  20. Investigating the educational effectiveness of a science museum exhibit on small modular fusion reactors

    E-Print Network [OSTI]

    Batie, Margo Alexandra

    2014-01-01

    Most people are unaware of the tremendous potential fusion reactors and smaller, more modular reactors possess. To inform them, a science exhibit was.constructed to investigate whether or not it would more effectively teach ...

  1. Slurry reactor design studies

    SciTech Connect (OSTI)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  2. Toward Infusing Modular and Reflective Design Learning throughout the Curriculum

    E-Print Network [OSTI]

    Georgas, John

    Toward Infusing Modular and Reflective Design Learning throughout the Curriculum John C. Georgas intervention that cen- ters on the widespread infusion of design learning throughout the curriculum using: An emphasis on broadly infusing design learning through the curriculum using modular design challenges

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

    SciTech Connect (OSTI)

    Johanna Oxstrand; Katya Le Blanc

    2014-07-01

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

  4. Johnson Noise Thermometry for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.; Holcomb, D.E.; Wood, R.T.

    2012-09-15

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

  5. Johnson Noise Thermometry for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Britton Jr, Charles L; Roberts, Michael; Bull, Nora D; Holcomb, David Eugene; Wood, Richard Thomas

    2012-10-01

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

  6. Evaluation of the Gas Turbine Modular Helium Reactor

    SciTech Connect (OSTI)

    Not Available

    1994-02-01

    Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs.

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

    SciTech Connect (OSTI)

    Loflin, Leonard; McRimmon, Beth

    2014-12-18

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

  8. Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

    SciTech Connect (OSTI)

    OHara J. M.; Higgins, J.; DAgostino, A.

    2012-01-17

    The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a single operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.

  9. Update on Small Modular Reactors Dynamics System Modeling Tool -- Molten Salt Cooled Architecture

    SciTech Connect (OSTI)

    Hale, Richard Edward; Cetiner, Sacit M.; Fugate, David L.; Qualls, A L.; Borum, Robert C.; Chaleff, Ethan S.; Rogerson, Doug W.; Batteh, John J.; Tiller, Michael M.

    2014-08-01

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

  10. Master thesis Modular mechanical engineering design tool (30hp) Background

    E-Print Network [OSTI]

    Zhao, Yuxiao

    Master thesis ­ Modular mechanical engineering design tool (30hp. Thesis description In this thesis, the students have the opportunity to develop-time. The thesis will be run in parallel with a customer order on a customized version

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

    SciTech Connect (OSTI)

    Curtis Smith

    2013-09-01

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

  12. A Framework to Expand and Advance Probabilistic Risk Assessment to Support Small Modular Reactors

    SciTech Connect (OSTI)

    Curtis Smith; David Schwieder; Robert Nourgaliev; Cherie Phelan; Diego Mandelli; Kellie Kvarfordt; Robert Youngblood

    2012-09-01

    During the early development of nuclear power plants, researchers and engineers focused on many aspects of plant operation, two of which were getting the newly-found technology to work and minimizing the likelihood of perceived accidents through redundancy and diversity. As time, and our experience, has progressed, the realization of plant operational risk/reliability has entered into the design, operation, and regulation of these plants. But, to date, we have only dabbled at the surface of risk and reliability technologies. For the next generation of small modular reactors (SMRs), it is imperative that these technologies evolve into an accepted, encompassing, validated, and integral part of the plant in order to reduce costs and to demonstrate safe operation. Further, while it is presumed that safety margins are substantial for proposed SMR designs, the depiction and demonstration of these margins needs to be better understood in order to optimize the licensing process.

  13. Small Modular Reactor: First of a Kind (FOAK) and Nth of a Kind (NOAK) Economic Analysis

    SciTech Connect (OSTI)

    Lauren M. Boldon; Piyush Sabharwall

    2014-08-01

    Small modular reactors (SMRs) refer to any reactor design in which the electricity generated is less than 300 MWe. Often medium sized reactors with power less than 700 MWe are also grouped into this category. Internationally, the development of a variety of designs for SMRs is booming with many designs approaching maturity and even in or nearing the licensing stage. It is for this reason that a generalized yet comprehensive economic model for first of a kind (FOAK) through nth of a kind (NOAK) SMRs based upon rated power, plant configuration, and the fiscal environment was developed. In the model, a particular project’s feasibility is assessed with regards to market conditions and by commonly utilized capital budgeting techniques, such as the net present value (NPV), internal rate of return (IRR), Payback, and more importantly, the levelized cost of energy (LCOE) for comparison to other energy production technologies. Finally, a sensitivity analysis was performed to determine the effects of changing debt, equity, interest rate, and conditions on the LCOE. The economic model is primarily applied to the near future water cooled SMR designs in the United States. Other gas cooled and liquid metal cooled SMR designs have been briefly outlined in terms of how the economic model would change. FOAK and NOAK SMR costs were determined for a site containing seven 180 MWe water cooled SMRs and compared to a site containing one 1260 MWe reactor. With an equal share of debt and equity and a 10% cost of debt and equity, the LCOE was determined to be $79 $84/MWh and $80/MWh for the SMR and large reactor sites, respectively. With a cost of equity of 15%, the SMR LCOE increased substantially to $103 $109/MWh. Finally, an increase in the equity share to 70% at the 15% cost of equity resulted in an even higher LCOE, demonstrating the large variation in results due to financial and market factors. The NPV and IRR both decreased with increasing LCOE. Unless the price of electricity increases along with the LCOE, the projects may become unprofitable. This is the case at the LCOE of $103 $109/MW, in which the NPV became negative. The IRR increased with increasing electricity price. Three cases, electric only base, storage—compressed air energy storage or pumped hydro, and hydrogen production, were performed incorporating SMRs into a nuclear wind natural gas hybrid energy system for the New York West Central region. The operational costs for three cases were calculated as $27/MWh, $25/MWh, and $28/MWh, respectively. A 3% increase in profits was demonstrated for the storage case over the electric only base case.

  14. METHODOLOGY Open Access Modular assembly of designer PUF proteins for

    E-Print Network [OSTI]

    Zhao, Huimin

    METHODOLOGY Open Access Modular assembly of designer PUF proteins for specific post of the PUF domain assembly method for RBP engineering, we fused the PUF domain to a post and applied biology and medicine. Keywords: Protein engineering, RNA-binding protein, Post

  15. SASSI Methodology-Based Sensitivity Studies for Deeply Embedded Structures, Such As Small Modular Reactors (SMRs)

    Broader source: Energy.gov [DOE]

    SASSI Methodology-Based Sensitivity Studies for Deeply Embedded Structures, Such As Small Modular Reactors (SMRs) Dr. Dan M. Ghiocel Ghiocel Predictive Technologies Inc. http://www.ghiocel-tech.com 2014 DOE Natural Phenomena Hazards Meeting Germantown, MD, October 21-22, 2014

  16. Technical Needs for Enhancing Risk Monitors with Equipment Condition Assessment for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Coles, Garill A.; Ramuhalli, Pradeep; Meyer, Ryan M.; Berglin, Eric J.; Wootan, David W.; Mitchell, Mark R.

    2013-04-04

    Advanced small modular reactors (aSMRs) can provide the United States with a safe, sustainable, and carbon-neutral energy source. The controllable day-to-day costs of aSMRs are expected to be dominated by operation and maintenance costs. Health and condition assessment coupled with online risk monitors can potentially enhance affordability of aSMRs through optimized operational planning and maintenance scheduling. Currently deployed risk monitors are an extension of probabilistic risk assessment (PRA). For complex engineered systems like nuclear power plants, PRA systematically combines event likelihoods and the probability of failure (POF) of key components, so that when combined with the magnitude of possible adverse consequences to determine risk. Traditional PRA uses population-based POF information to estimate the average plant risk over time. Currently, most nuclear power plants have a PRA that reflects the as-operated, as-modified plant; this model is updated periodically, typically once a year. Risk monitors expand on living PRA by incorporating changes in the day-by-day plant operation and configuration (e.g., changes in equipment availability, operating regime, environmental conditions). However, population-based POF (or population- and time-based POF) is still used to populate fault trees. Health monitoring techniques can be used to establish condition indicators and monitoring capabilities that indicate the component-specific POF at a desired point in time (or over a desired period), which can then be incorporated in the risk monitor to provide a more accurate estimate of the plant risk in different configurations. This is particularly important for active systems, structures, and components (SSCs) proposed for use in aSMR designs. These SSCs may differ significantly from those used in the operating fleet of light-water reactors (or even in LWR-based SMR designs). Additionally, the operating characteristics of aSMRs can present significantly different requirements, including the need to operate in different coolant environments, higher operating temperatures, and longer operating cycles between planned refueling and maintenance outages. These features, along with the relative lack of operating experience for some of the proposed advanced designs, may limit the ability to estimate event probability and component POF with a high degree of certainty. Incorporating real-time estimates of component POF may compensate for a relative lack of established knowledge about the long-term component behavior and improve operational and maintenance planning and optimization. The particular eccentricities of advanced reactors and small modular reactors provide unique challenges and needs for advanced instrumentation, control, and human-machine interface (ICHMI) techniques such as enhanced risk monitors (ERM) in aSMRs. Several features of aSMR designs increase the need for accurate characterization of the real-time risk during operation and maintenance activities. A number of technical gaps in realizing ERM exist, and these gaps are largely independent of the specific reactor technology. As a result, the development of a framework for ERM would enable greater situational awareness regardless of the specific class of reactor technology. A set of research tasks are identified in a preliminary research plan to enable the development, testing, and demonstration of such a framework. Although some aspects of aSMRs, such as specific operational characteristics, will vary and are not now completely defined, the proposed framework is expected to be relevant regardless of such uncertainty. The development of an ERM framework will provide one of the key technical developments necessary to ensure the economic viability of aSMRs.

  17. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    SciTech Connect (OSTI)

    Sonat Sen; Gilles Youinou

    2013-02-01

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  18. Small Modular Reactors Presentation to Secretary of Energy Advisory...

    Energy Savers [EERE]

    and demonstration of innovative reactor technologies and concepts JohnKelly-SEABSMRBriefingJuly202011final.pdf More Documents & Publications Before the Senate...

  19. Modular hybrid plasma reactor and related systems and methods

    DOE Patents [OSTI]

    Kong, Peter C.; Grandy, Jon D.; Detering, Brent A.

    2010-06-22

    A device, method and system for generating a plasma is disclosed wherein an electrical arc is established and the movement of the electrical arc is selectively controlled. In one example, modular units are coupled to one another to collectively define a chamber. Each modular unit may include an electrode and a cathode spaced apart and configured to generate an arc therebetween. A device, such as a magnetic or electromagnetic device, may be used to selectively control the movement of the arc about a longitudinal axis of the chamber. The arcs of individual modules may be individually controlled so as to exhibit similar or dissimilar motions about the longitudinal axis of the chamber. In another embodiment, an inlet structure may be used to selectively define the flow path of matter introduced into the chamber such that it travels in a substantially circular or helical path within the chamber.

  20. NRC Reviewer Aid for Evaluating the Human Factors Engineering Aspects of Small Modular Reactors

    SciTech Connect (OSTI)

    OHara J. M.; Higgins, J.C.

    2012-01-13

    Small modular reactors (SMRs) are a promising approach to meeting future energy needs. Although the electrical output of an individual SMR is relatively small compared to that of typical commercial nuclear plants, they can be grouped to produce as much energy as a utility demands. Furthermore, SMRs can be used for other purposes, such as producing hydrogen and generating process heat. The design characteristics of many SMRs differ from those of current conventional plants and may require a distinct concept of operations (ConOps). The U.S. Nuclear Regulatory Commission (NRC) conducted research to examine the human factors engineering (HFE) and the operational aspects of SMRs. The research identified thirty potential human-performance issues that should be considered in the NRC's reviews of SMR designs and in future research activities. The purpose of this report is to support NRC HFE reviewers of SMR applications by identifying some of the questions that can be asked of applicants whose designs have characteristics identified in the issues. The questions for each issue were identified and organized based on the review elements and guidance contained in Chapter 18 of the Standard Review Plan (NUREG-0800), and the Human Factors Engineering Program Review Model (NUREG-0711).

  1. Progress Towards Prognostic Health Management of Passive Components in Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Ramuhalli, Pradeep; Hirt, Evelyn H.; Pardini, Allan F.; Suter, Jonathan D.; Prowant, Matthew S.

    2014-08-01

    Sustainable nuclear power to promote energy security and to reduce greenhouse gas emissions are two key national energy priorities. The development of deployable small modular reactors (SMRs) is expected to support these objectives by developing technologies that improve the reliability, sustain safety, and improve affordability of new reactors. Advanced SMRs (AdvSMRs) refer to a specific class of SMRs and are based on modularization of advanced reactor concepts. Prognostic health management (PHM) systems can benefit both the safety and economics of deploying AdvSMRs and can play an essential role in managing the inspection and maintenance of passive components in AdvSMR systems. This paper describes progress on development of a prototypic PHM system for AdvSMR passive components, with thermal creep chosen as the target degradation mechanism.

  2. Incorporating Equipment Condition Assessment in Risk Monitors for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Coles, Garill A.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-10-01

    Advanced small modular reactors (aSMRs) can complement the current fleet of large light-water reactors in the USA for baseload and peak demand power production and process heat applications (e.g., water desalination, shale oil extraction, hydrogen production). The day-to-day costs of aSMRs are expected to be dominated by operations and maintenance (O&M); however, the effect of diverse operating missions and unit modularity on O&M is not fully understood. These costs could potentially be reduced by optimized scheduling, with risk-informed scheduling of maintenance, repair, and replacement of equipment. Currently, most nuclear power plants have a “living” probabilistic risk assessment (PRA), which reflects the as-operated, as-modified plant and combine event probabilities with population-based probability of failure (POF) for key components. “Risk monitors” extend the PRA by incorporating the actual and dynamic plant configuration (equipment availability, operating regime, environmental conditions, etc.) into risk assessment. In fact, PRAs are more integrated into plant management in today’s nuclear power plants than at any other time in the history of nuclear power. However, population-based POF curves are still used to populate fault trees; this approach neglects the time-varying condition of equipment that is relied on during standard and non-standard configurations. Equipment condition monitoring techniques can be used to estimate the component POF. Incorporating this unit-specific estimate of POF in the risk monitor can provide a more accurate estimate of risk in different operating and maintenance configurations. This enhanced risk assessment will be especially important for aSMRs that have advanced component designs, which don’t have an available operating history to draw from, and often use passive design features, which present challenges to PRA. This paper presents the requirements and technical gaps for developing a framework to integrate unit-specific estimates of POF into risk monitors, resulting in enhanced risk monitors that support optimized operation and maintenance of aSMRs.

  3. MIT Modular Pebble Bed Reactor (MPBR) A Summary of Research Activities and Accomplishments

    E-Print Network [OSTI]

    Lanza · Martin Bazant (Math) #12;#12;#12;#12;Our Vision for 1150 MW Combined Heat and Power Station Air Ingress · Balance of Plant Design · Modularity Design · Intermediate Heat Exchanger Design · Core Power Distribution Monitoring · Pebble Flow Experiments · Non-Proliferation · Safeguards · Waste

  4. Westinghouse Small Modular Reactor passive safety system response to postulated events

    SciTech Connect (OSTI)

    Smith, M. C.; Wright, R. F. [Westinghouse Electric Company, 600 Cranberry Woods Drive (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to reduce the reactor pressure in a controlled manner to facilitate the passive injection. Long-term decay heat removal is accomplished using the passive heat removal systems augmented by heat transfer through the containment vessel to the environment. The passive injection systems are designed so that the fuel remains covered and effectively cooled throughout the event. Like during the frequent faults, the passive systems provide effective cooling without the need for ac power for seven days following the accident. Connections are available to add additional water to indefinitely cool the plant. The response of the safety systems of the Westinghouse SMR to various initiating faults has been examined. Among them, two accidents; an extended station blackout event, and a LOCA event have been evaluated to demonstrate how the plant will remain safe in the unlikely event that either should occur. (authors)

  5. Population Sensitivity Evaluation of Two Proposed Hampton Roads Area Sites for a Possible Small Modular Reactor

    SciTech Connect (OSTI)

    Belles, R. J.; Omitaomu, O. A.

    2014-08-01

    The overall objective of this research project is to use the OR-SAGE tool to support the US Department of Energy (DOE) Office of Nuclear Energy (NE) in evaluating future electrical generation deployment options for small modular reactors (SMRs) in areas with significant energy demand from the federal sector. Deployment of SMRs in zones with high federal energy use can provide a means of meeting federal clean energy goals.

  6. Small Modular Reactor Report (SEAB) | Department of Energy

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

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirley Ann JacksonDepartment ofOffice|inWestMayBuilding K-25 cleanup at theProjects |Reactor

  7. CONFERENCES AND SYMPOSIA FUSION REACTOR DESIGN IV

    E-Print Network [OSTI]

    Abdou, Mohamed

    in the physics of laser-target interactions, target design and implosion experiments; 5.3. New ICF reactorCONFERENCES AND SYMPOSIA FUSION REACTOR DESIGN IV Report on the Fourth IAEA Technical Committee Reactor Design and Technology at Yalta, USSR, from 26 May -- 6 June 1986. This report contains all

  8. Nuclear Reactor Safety Design Criteria

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

    1993-01-19

    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. Supersedes DOE 5480.1, dated 1-19-93. Certified 11-18-10.

  9. Modular Hybrid Plasma Reactor for Low Cost Bulk Production of Nanomaterials

    SciTech Connect (OSTI)

    Peter C. Kong

    2011-12-01

    INL developed a bench scale modular hybrid plasma system for gas phase nanomaterials synthesis. The system was being optimized for WO3 nanoparticles production and scale model projection to a 300 kW pilot system. During the course of technology development many modifications had been done to the system to resolve technical issues that had surfaced and also to improve the performance. All project tasks had been completed except 2 optimization subtasks. These 2 subtasks, a 4-hour and an 8-hour continuous powder production runs at 1 lb/hr powder feeding rate, were unable to complete due to technical issues developed with the reactor system. The 4-hour run had been attempted twice and both times the run was terminated prematurely. The modular electrode for the plasma system was significantly redesigned to address the technical issues. Fabrication of the redesigned modular electrodes and additional components had been completed at the end of the project life. However, not enough resource was available to perform tests to evaluate the performance of the new modifications. More development work would be needed to resolve these problems prior to scaling. The technology demonstrated a surprising capability of synthesizing a single phase of meta-stable delta-Al2O3 from pure alpha-phase large Al2O3 powder. The formation of delta-Al2O3 was surprising because this phase is meta-stable and only formed between 973-1073 K, and delta-Al2O3 is very difficult to synthesize as a single phase. Besides the specific temperature window to form this phase, this meta-stable phase may have been stabilized by nanoparticle size formed in a high temperature plasma process. This technology may possess the capability to produce unusual meta-stable nanophase materials that would be otherwise difficult to produce by conventional methods. A 300 kW INL modular hybrid plasma pilot scale model reactor had been projected using the experimental data from PPG Industries 300 kW hot wall plasma reactor. The projected size of the INL 300 kW pilot model reactor would be about 15% that of the PPG 300 kW hot wall plasma reactor. Including the safety net factor the projected INL pilot reactor size would be 25-30% of the PPG 300 kW hot wall plasma pilot reactor. Due to the modularity of the INL plasma reactor and the energy cascading effect from the upstream plasma to the downstream plasma the energy utilization is more efficient in material processing. It is envisioning that the material through put range for the INL pilot reactor would be comparable to the PPG 300 kW pilot reactor but the energy consumption would be lower. The INL hybrid plasma technology is rather close to being optimized for scaling to a pilot system. More near term development work is still needed to complete the process optimization before pilot scaling.

  10. Human-System Interfaces (HSIs) in Small Modular Reactors (SMRs)

    SciTech Connect (OSTI)

    Jacques V Hugo

    2014-10-01

    This book chapter describes the considerations for the selection of advanced human–system interfaces (HSIs) for the new generation of nuclear power plants. The chapter discusses the technologies that will be needed to support highly automated nuclear power plants, while minimising demands for numbers of operational staff, reducing human error and improving plant efficiency and safety. Special attention is paid to the selection and deployment of advanced technologies in nuclear power plants (NPPs). The chapter closes with an examination of how technologies are likely to develop over the next 10–15 years and how this will affect design choices for the nuclear industry.

  11. U.S. Department of Energy Instrumentation and Controls Technology Research for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Wood, Richard Thomas

    2012-01-01

    Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD&D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, key DOE programs have substantial ICHMI RD&D elements to their respective research portfolio. This article describes current ICHMI research to support the development of advanced small modular reactors.

  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.

    2003-11-12

    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 and Evaluation of a Modular Resonant Switched Capacitors Equalizer for PV Panels

    E-Print Network [OSTI]

    Design and Evaluation of a Modular Resonant Switched Capacitors Equalizer for PV Panels Shmuel (Sam of shaded panels in a serially connected PV array. The proposed solution is based on a modular approach module was designed for 185W PV panels and was found to boost the maximum available power by about 50

  14. INITIATORS AND TRIGGERING CONDITIONS FOR ADAPTIVE AUTOMATION IN ADVANCED SMALL MODULAR REACTORS

    SciTech Connect (OSTI)

    Katya L Le Blanc; Johanna h Oxstrand

    2014-04-01

    It is anticipated that Advanced Small Modular Reactors (AdvSMRs) will employ high degrees of automation. High levels of automation can enhance system performance, but often at the cost of reduced human performance. Automation can lead to human out-of the loop issues, unbalanced workload, complacency, and other problems if it is not designed properly. Researchers have proposed adaptive automation (defined as dynamic or flexible allocation of functions) as a way to get the benefits of higher levels of automation without the human performance costs. Adaptive automation has the potential to balance operator workload and enhance operator situation awareness by allocating functions to the operators in a way that is sensitive to overall workload and capabilities at the time of operation. However, there still a number of questions regarding how to effectively design adaptive automation to achieve that potential. One of those questions is related to how to initiate (or trigger) a shift in automation in order to provide maximal sensitivity to operator needs without introducing undesirable consequences (such as unpredictable mode changes). Several triggering mechanisms for shifts in adaptive automation have been proposed including: operator initiated, critical events, performance-based, physiological measurement, model-based, and hybrid methods. As part of a larger project to develop design guidance for human-automation collaboration in AdvSMRs, researchers at Idaho National Laboratory have investigated the effectiveness and applicability of each of these triggering mechanisms in the context of AdvSMR. Researchers reviewed the empirical literature on adaptive automation and assessed each triggering mechanism based on the human-system performance consequences of employing that mechanism. Researchers also assessed the practicality and feasibility of using the mechanism in the context of an AdvSMR control room. Results indicate that there are tradeoffs associated with each mechanism, but that some are more applicable to the AdvSMR domain. The two mechanisms that consistently improve performance in laboratory studies are operator initiated adaptive automation based on hierarchical task delegation and the Electroencephalogram(EEG) –based measure of engagement. Current EEG methods are intrusive and require intensive analysis; therefore it is not recommended for an AdvSMR control rooms at this time. Researchers also discuss limitations in the existing empirical literature and make recommendations for further research.

  15. Assessment of passive decay heat removal in the General Atomics Modular Helium Reactor 

    E-Print Network [OSTI]

    Cocheme, Francois Guilhem

    2005-02-17

    /ATHENA. The MHR is a high temperature gas cooled reactor. It is a prismatic core concept for New Generation Nuclear Plant (NGNP). Very few reactors of that kind have been designed in the past. Furthermore, the MHR is supposed to be a highly passively safe concept...

  16. Reactor protection system design alternatives for sodium fast reactors

    E-Print Network [OSTI]

    DeWitte, Jacob D. (Jacob Dominic)

    2011-01-01

    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 ...

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

    E-Print Network [OSTI]

    Pope, Michael A. (Michael Alexander)

    2004-01-01

    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 ...

  18. Conceptual designs for modular OTEC SKSS. Final report

    SciTech Connect (OSTI)

    None

    1980-02-29

    This volume presents the results of the first phase of the Station Keeping Subsystem (SKSS) design study for 40 MW/sub e/ capacity Modular Experiment OTEC Platforms. The objectives of the study were: (1) establishment of basic design requirements; (2) verification of technical feasibility of SKSS designs; (3) identification of merits and demerits; (4) estimates of sizes for major components; (5) estimates of life cycle costs; (6) deployment scenarios and time/cost/risk assessments; (7) maintenance/repair and replacement scenarios; (8) identifications of interface with other OTEC subsystems; (9) recommendations for and major problems in preliminary design; and (10) applicability of concepts to commercial plant SKSS designs. A brief site suitability study was performed with the objective of determining the best possible location at the Punta Tuna (Puerto Rico) site from the standpoint of anchoring. This involved studying the vicinity of the initial location in relation to the prevailing bottom slopes and distances from shore. All subsequent studies were performed for the final selected site. The two baseline OTEC platforms were the APL BARGE and the G and C SPAR. The results of the study are presented in detail. The overall objective of developing two conceptual designs for each of the two baseline OTEC platforms has been accomplished. Specifically: (1) a methodology was developed for conceptual designs and followed to the extent possible. At this stage, a full reliability/performance/optimization analysis based on a probabilistic approach was not used due to the numerous SKSS candidates to be evaluated. A deterministic approach was used. (2) For both of the two baseline platforms, the APL BARGE and the G and C SPAR, all possible SKSS candidate concepts were considered and matrices of SKSS concepts were developed.

  19. Modular Coil Design for the Ultra-low Aspect Ratio Quasi-axially Symmetric Stellarator MHH2

    SciTech Connect (OSTI)

    Ku LP, the ARIES-CS Team

    2005-09-27

    A family of two field-period quasi-axisymmetric stellarators generally known as MHH2 with aspect ratios of only {approx}2.5 was found. These configurations have low field ripples and excellent confinement of {alpha} particles. This discovery raises the hope that a compact stellarator reactor may eventually be designed with the property of tokamak transport and stellarator stability. In this paper we demonstrate that smooth modular coils may be designed for this family of configurations that not only yield plasmas with good physics properties but also possess engineering properties desirable for compact power producing reactors. We show designs featuring 16 modular coils with ratios of major radius to minimum coil-plasma separation {approx}5.5, major radius to minimum coil-coil separation {approx}10 and the maximum field in coil bodies to the field on axis {approx}2 for 0.2 m{sup 2} conductors. These coils is expected to allow plasmas operated at 5% {beta} with {alpha} energy loss < 10% for a reactor of major radius <9 m at 5 T.

  20. Development of the Mathematics of Learning Curve Models for Evaluating Small Modular Reactor Economics

    SciTech Connect (OSTI)

    Harrison, T. J. [ORNL

    2014-02-01

    The cost of nuclear power is a straightforward yet complicated topic. It is straightforward in that the cost of nuclear power is a function of the cost to build the nuclear power plant, the cost to operate and maintain it, and the cost to provide fuel for it. It is complicated in that some of those costs are not necessarily known, introducing uncertainty into the analysis. For large light water reactor (LWR)-based nuclear power plants, the uncertainty is mainly contained within the cost of construction. The typical costs of operations and maintenance (O&M), as well as fuel, are well known based on the current fleet of LWRs. However, the last currently operating reactor to come online was Watts Bar 1 in May 1996; thus, the expected construction costs for gigawatt (GW)-class reactors in the United States are based on information nearly two decades old. Extrapolating construction, O&M, and fuel costs from GW-class LWRs to LWR-based small modular reactors (SMRs) introduces even more complication. The per-installed-kilowatt construction costs for SMRs are likely to be higher than those for the GW-class reactors based on the property of the economy of scale. Generally speaking, the economy of scale is the tendency for overall costs to increase slower than the overall production capacity. For power plants, this means that doubling the power production capacity would be expected to cost less than twice as much. Applying this property in the opposite direction, halving the power production capacity would be expected to cost more than half as much. This can potentially make the SMRs less competitive in the electricity market against the GW-class reactors, as well as against other power sources such as natural gas and subsidized renewables. One factor that can potentially aid the SMRs in achieving economic competitiveness is an economy of numbers, as opposed to the economy of scale, associated with learning curves. The basic concept of the learning curve is that the more a new process is repeated, the more efficient the process can be made. Assuming that efficiency directly relates to cost means that the more a new process is repeated successfully and efficiently, the less costly the process can be made. This factor ties directly into the factory fabrication and modularization aspect of the SMR paradigm—manufacturing serial, standardized, identical components for use in nuclear power plants can allow the SMR industry to use the learning curves to predict and optimize deployment costs.

  1. Design and implementation of a relative state estimator for docking and formation control of modular autonomous spacecraft

    E-Print Network [OSTI]

    Hoff, Nicholas R

    2007-01-01

    Modularity is a promising design concept for space systems. In a modular satellite, the individual subsystems would be broken down into physically distinct modules, which would then dynamically recombine into an aggregate ...

  2. The Structure and Value of Modularity in Software Design Kevin Sullivan, Yuanfang Cai, Ben Hallen

    E-Print Network [OSTI]

    Huang, Wei

    The Structure and Value of Modularity in Software Design Kevin Sullivan, Yuanfang Cai, Ben Hallen design. The theory uses design structure matrices to model designs and real options techniques to value in its terms--Parnas's KWIC--and evaluate the results. We contribute an extension to design structure

  3. Design of electronics for a high-resolution, multi-material, and modular 3D printer

    E-Print Network [OSTI]

    Kwan, Joyce G

    2013-01-01

    Electronics for a high-resolution, multi-material, and modular 3D printer were designed and implemented. The driver for a piezoelectric inkjet print head can fire its nozzles with one of three droplet sizes ranging from 6 ...

  4. Design, Control and Motion Planning for a Novel Modular Extendable Robotic Manipulator 

    E-Print Network [OSTI]

    Yi, Hak 1979-

    2012-12-05

    This dissertation discusses an implementation of a design, control and motion planning for a novel extendable modular redundant robotic manipulator in space constraints, which robots may encounter for completing required tasks in small...

  5. Design and analysis of a concrete modular housing system constructed with 3D panels

    E-Print Network [OSTI]

    Sarcia, Sam Rhea, 1982-

    2004-01-01

    An innovative modular house system design utilizing an alternative concrete residential building system called 3D panels is presented along with an overview of 3D panels as well as relevant methods and markets. The proposed ...

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

    SciTech Connect (OSTI)

    Trond Bjornard; John Hockert

    2011-08-01

    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.

  7. Evaluation of the applicability of existing nuclear power plant regulatory requirements in the U.S. to advanced small modular reactors.

    SciTech Connect (OSTI)

    LaChance, Jeffrey L.; Wheeler, Timothy A.; Farnum, Cathy Ottinger; Middleton, Bobby D.; Jordan, Sabina Erteza; Duran, Felicia Angelica; Baum, Gregory A.

    2013-05-01

    The current wave of small modular reactor (SMR) designs all have the goal of reducing the cost of management and operations. By optimizing the system, the goal is to make these power plants safer, cheaper to operate and maintain, and more secure. In particular, the reduction in plant staffing can result in significant cost savings. The introduction of advanced reactor designs and increased use of advanced automation technologies in existing nuclear power plants will likely change the roles, responsibilities, composition, and size of the crews required to control plant operations. Similarly, certain security staffing requirements for traditional operational nuclear power plants may not be appropriate or necessary for SMRs due to the simpler, safer and more automated design characteristics of SMRs. As a first step in a process to identify where regulatory requirements may be met with reduced staffing and therefore lower cost, this report identifies the regulatory requirements and associated guidance utilized in the licensing of existing reactors. The potential applicability of these regulations to advanced SMR designs is identified taking into account the unique features of these types of reactors.

  8. The Structure and Value of Modularity in Software Design Kevin J. Sullivan

    E-Print Network [OSTI]

    Weimer, Westley

    The Structure and Value of Modularity in Software Design Kevin J. Sullivan University of Virginia uses design structure matrices to model designs and real options techniques to value them. To test an extension to design structure matrices, and we show that the options results are consistent with Parnas

  9. Reactor Technology | Nuclear Science | ORNL

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

    Reactor Technology Advanced Reactor Concepts Advanced Instrumentation & Controls Light Water Reactor Sustainability Safety and Regulatory Technology Small Modular Reactors Nuclear...

  10. Advances in ICF power reactor design

    SciTech Connect (OSTI)

    Hogan, W.J.; Kulcinski, G.L.

    1985-07-01

    Fifteen ICF power reactor design studies published since 1980 are reviewed to illuminate the design trends they represent. There is a clear, continuing trend toward making ICF reactors inherently safer and environmentally benign. Since this trend accentuates inherent advantages of ICF reactors, we expect it to be further emphasized in the future. An emphasis on economic competitiveness appears to be a somewhat newer trend. Lower cost of electricity, smaller initial size (and capital cost), and more affordable development paths are three of the issues being addressed with new studies.

  11. Mirror Advanced Reactor Study interim design report

    SciTech Connect (OSTI)

    Not Available

    1983-04-01

    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.

  12. Optimal Topological Design of Infinite Modular Structures Moshe B. Fuchs and Michael Ryvkin

    E-Print Network [OSTI]

    Fuchs, Moshe

    Optimal Topological Design of Infinite Modular Structures Moshe B. Fuchs and Michael Ryvkin Tel Aviv University 1. Abstract This paper deals with the optimum topological design of periodic structures are in wide use but have not received much attention in structural analysis, let alone structural design

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

    SciTech Connect (OSTI)

    Anheier, Norman C.; Suter, Jonathan D.; Qiao, Hong; Andersen, Eric S.; Berglin, Eric J.; Bliss, Mary; Cannon, Bret D.; Devanathan, Ramaswami; Mendoza, Albert; Sheen, David M.

    2013-08-06

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

  14. Identification of Selected Areas to Support Federal Clean Energy Goals Using Small Modular Reactors

    SciTech Connect (OSTI)

    Belles, Randy; Mays, Gary T; Omitaomu, Olufemi A; Poore III, Willis P

    2013-12-01

    This analysis identifies candidate locations, in a broad sense, where there are high concentrations of federal government agency use of electricity, which are also suitable areas for near-term SMRs. Near-term SMRs are based on light-water reactor (LWR) technology with compact design features that are expected to offer a host of safety, siting, construction, and economic benefits. These smaller plants are ideally suited for small electric grids and for locations that cannot support large reactors, thus providing utilities or governement entities with the flexibility to scale power production as demand changes by adding additional power by deploying more modules or reactors in phases. This research project is aimed at providing methodologies, information, and insights to assist the federal government in meeting federal clean energy goals.

  15. Design options for a bunsen reactor.

    SciTech Connect (OSTI)

    Moore, Robert Charles

    2013-10-01

    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. Identification of Selected Areas to Support Federal Clean Energy Goals Using Small Modular Reactors

    SciTech Connect (OSTI)

    Belles, R. J. [ORNL; Mays, G. T. [ORNL; Omitaomu, O. A. [ORNL; Poore, W. P. [ORNL

    2013-12-30

    Beginning in late 2008, Oak Ridge National Laboratory (ORNL) responded to ongoing internal and external studies addressing key questions related to our national electrical energy supply. This effort has led to the development and refinement of Oak Ridge Siting Analysis for power Generation Expansion (OR-SAGE), a tool to support power plant siting evaluations. The objective in developing OR-SAGE was to use industry-accepted approaches and/or develop appropriate criteria for screening sites and employ an array of geographic information systems (GIS) data sources at ORNL to identify candidate areas for a power generation technology application. The basic premise requires the development of exclusionary, avoidance, and suitability criteria for evaluating sites for a given siting application, such as siting small modular reactors (SMRs). For specific applications of the tool, it is necessary to develop site selection and evaluation criteria (SSEC) that encompass a number of key benchmarks that essentially form the site environmental characterization for that application. These SSEC might include population density, seismic activity, proximity to water sources, proximity to hazardous facilities, avoidance of protected lands and floodplains, susceptibility to landslide hazards, and others.

  17. Cost-Shared Development of Innovative Small Modular Reactor Designs |

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

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyTher i n c i p aDepartmentEnergy comparingDeepDecemberCornstalksAmerica

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

    SciTech Connect (OSTI)

    John Darrell Bess

    2008-06-01

    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.

  19. Modular Dual Coolant Pb-17Li Blanket Design For ARIES-CS Compact Stellarator Power Plant

    E-Print Network [OSTI]

    Raffray, A. René

    of the study. The preferred blanket concept is a dual coolant blanket with a He- cooled ferritic steel firstModular Dual Coolant Pb-17Li Blanket Design For ARIES-CS Compact Stellarator Power Plant X.R. Wanga from the engineering effort during the second phase of ARIES-CS study on the conceptual design

  20. Designing Urban Green Roofs for Modularity and Recyclability Objective: Develop alternative designs for green roofs responsive to the special

    E-Print Network [OSTI]

    Wolberg, George

    a significant payoff in reduced heat loss and higher energy efficiency. However, current green roof proposals, energy supply systems, and monitoring systems. From an architectural perspective, develop schematic greenDesigning Urban Green Roofs for Modularity and Recyclability Objective: Develop alternative designs

  1. Stimuli-Responsive Smart Gels Realized via Modular Protein Design Tijana Z. Grove,

    E-Print Network [OSTI]

    Regan, Lynne

    Stimuli-Responsive Smart Gels Realized via Modular Protein Design Tijana Z. Grove, Chinedum O 26, 2010; E-mail: lynne.regan@yale.edu Abstract: Smart gels have a variety of applications, including that permits the creation of protein-based smart gels with encoded morphology, functionality

  2. Designing Modular Hardware Accelerators in C With ROCCC 2.0

    E-Print Network [OSTI]

    Najjar, Walid A.

    Designing Modular Hardware Accelerators in C With ROCCC 2.0 Jason Villarreal, Adrian Park Jacquard, Riverside {najjar, rhalstea}@cs.ucr.edu Abstract--While FPGA-based hardware accelerators have re- peatedly acceptance by application code developers. These platforms are typically programmed in a low level hardware

  3. 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

    2008-12-16

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

  4. U.S. Department Of Energy Advanced Small Modular Reactor R&D Program: Instrumentation, Controls, and Human-Machine Interface (ICHMI) Pathway

    SciTech Connect (OSTI)

    Holcomb, David Eugene; Wood, Richard Thomas

    2013-01-01

    Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The nuclear power industry is currently engaged in a transition from traditional analog-based instrumentation, controls, and human-machine interface systems to implementations employing digital technologies. This transition has primarily occurred in an ad hoc fashion through individual system upgrades at existing plants and has been constrained by licenseability concerns. Although the recent progress in constructing new plants has spurred design of more fully digital plant-wide ICHMI systems, the experience base in the nuclear power application domain is limited. Additionally, development of advanced reactor concepts, such as Generation IV designs and small modular reactors, introduces different plant conditions (e.g., higher temperatures, different coolants, etc.) and unique plant configurations (e.g., multiunit plants with shared systems, balance of plant architectures with reconfigurable co-generation options) that increase the need for enhanced ICHMI capabilities to fully achieve industry goals related to economic competitiveness, safety and reliability, sustainability, and proliferation resistance and physical protection. As a result, significant challenges remain to be addressed to enable the nuclear power industry to complete the transition to safe and comprehensive use of modern ICHMI technology. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD&D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, several DOE programs have substantial ICHMI RD&D elements within their respective research portfolios. This paper describes current ICHMI research in support of advanced small modular reactors. The objectives that can be achieved through execution of the defined RD&D are to provide optimal technical solutions to critical ICHMI issues, resolve technology gaps arising from the unique measurement and control characteristics of advanced reactor concepts, provide demonstration of needed technologies and methodologies in the nuclear power application domain, mature emerging technologies to facilitate commercialization, and establish necessary technical evidence and application experience to enable timely and predictable licensing. 1 Introduction Instrumentation, controls, and human-machine interfaces are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The nuclear power industry is currently engaged in a transition from traditional analog-based instrumentation, controls, and human-machine interface (ICHMI) systems to implementations employing digital technologies. This transition has primarily occurred in an ad hoc fashion through individual system upgrades at existing plants and has been constrained by licenseability concerns. Although the recent progress in constructing new plants has spurred design of more fully digital plant-wide ICHMI systems, the experience base in the nuclear power application domain is limited. Additionally, development of advanced reactor concepts, such as Generation IV designs and small modular reactors, introduces different plant conditions (e.g., higher temperatures, different coolants, etc.) and unique plant configurations (e.g., multiunit plants with shared systems, balance of plant architectures with reconfigurable co-generation options) that increase the need for enhanced ICHMI capabilities to fully achieve industry goals related to economic competitiveness, safety and reliability, sustainability, and proliferation resistance and physical protection. As a result, significant challenges remain to be addressed to enable the nuclear power industry to complete the transition to safe and comprehensive use of m

  5. Conceptual Design of a 100 MWe Modular Molten Salt Power Tower Plant

    SciTech Connect (OSTI)

    James E. Pacheco; Carter Moursund, Dale Rogers, David Wasyluk

    2011-09-20

    A conceptual design of a 100 MWe modular molten salt solar power tower plant has been developed which can provide capacity factors in the range of 35 to 75%. Compared to single tower plants, the modular design provides a higher degree of flexibility in achieving the desired customer's capacity factor and is obtained simply by adjusting the number of standard modules. Each module consists of a standard size heliostat field and receiver system, hence reengineering and associated unacceptable performance uncertainties due to scaling are eliminated. The modular approach with multiple towers also improves plant availability. Heliostat field components, receivers and towers are shop assembled allowing for high quality and minimal field assembly. A centralized thermal-storage system stores hot salt from the receivers, allowing nearly continuous power production, independent of solar energy collection, and improved parity with the grid. A molten salt steam generator converts the stored thermal energy into steam, which powers a steam turbine generator to produce electricity. This paper describes the conceptual design of the plant, the advantages of modularity, expected performance, pathways to cost reductions, and environmental impact.

  6. Innovative design of uranium startup fast reactors

    E-Print Network [OSTI]

    Fei, Tingzhou

    2012-01-01

    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 ...

  7. Measurement of Temperature Profile in the Reactor Cavity Cooling System 

    E-Print Network [OSTI]

    Alhashimi, Tariq Yaqoob Sayed

    2014-12-02

    -cooled Reactor Cavity Cooling System (RCCS) based on General Electric Modular High Temperature Gas Cooled Reactor (MHTGR) design to study the thermal hydraulic phenomena occurring in the upper plenum. The facility consists of four vertical parallel riser ducts...

  8. NRC policy on future reactor designs

    SciTech Connect (OSTI)

    none,

    1985-07-01

    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.

  9. Neutron transport analysis for nuclear reactor design

    DOE Patents [OSTI]

    Vujic, Jasmina L. (Lisle, IL)

    1993-01-01

    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.

  10. Neutron transport analysis for nuclear reactor design

    DOE Patents [OSTI]

    Vujic, J.L.

    1993-11-30

    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.

  11. Dependency models as a basis for analyzing software product platform modularity : a case study in strategic software design rationalization

    E-Print Network [OSTI]

    LaMantia, Matthew J. (Matthew John)

    2006-01-01

    It is broadly accepted among software managers and architects that maintaining the integrity of software designs is important for the long-term health and viability of software product platforms. The use of modular, ...

  12. Assessment of Algal Farm Designs using a Dynamic Modular Approach

    SciTech Connect (OSTI)

    Abodeely, Jared M. [Idaho National Laboratory (INL), Idaho Falls, ID (United States). Biofuels and Renewable Energy Technology; Stevens, Daniel M. [Idaho National Laboratory (INL), Idaho Falls, ID (United States). Biofuels and Renewable Energy Technology; Ray, Allison E. [Idaho National Laboratory (INL), Idaho Falls, ID (United States). Biofuels and Renewable Energy Technology; Newby, Deborah T. [Idaho National Laboratory (INL), Idaho Falls, ID (United States). Biofuels and Renewable Energy Technology; Coleman, Andre M. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States). Hydrology Technical Group; Cafferty, Kara G. [Idaho National Laboratory (INL), Idaho Falls, ID (United States). Biofuels and Renewable Energy Technology

    2014-07-01

    The notion of renewable energy provides an importantmechanism for diversifying an energy portfolio,which ultimately would have numerous benefits including increased energy resilience, reduced reliance on foreign energysupplies, reduced GHG emissions, development of a green energy sector that contributes to economic growth,and providing a sustainable energy supply. The conversion of autotrophic algae to liquid transportation fuels is the basis of several decades of research to competitively bring energy-scale production into reality; however, many challenges still remain for making algal biofuels economically viable. Addressing current challenges associatedwith algal production systems, in part, requires the ability to assess spatial and temporal variability, rapidly evaluate alternative algal production system designs, and perform large-scale assessments considering multiple scenarios for thousands of potential sites. We introduce the development and application of the Algae Logistics Model (ALM) which is tailored to help address these challenges. The flexible nature of the ALM architecture allows the model to: 1) interface with external biomass production and resource assessment models, as well as other relevant datasets including those with spatiotemporal granularity; 2) interchange design processes to enable operational and economic assessments ofmultiple design configurations, including the integration of current and new innovative technologies; and 3) conduct trade-off analysis to help understand the site-specific techno-economic trade-offs and inform technology decisions. This study uses the ALM to investigate a baseline open-pond production system determined by model harmonization efforts conducted by the U.S. Department of Energy. Six sites in the U.S. southern-tierwere sub-selected and assessed using daily site-specific algaebiomass productivity data to determine the economic viability of large-scale open-pond systems. Results show that costs can vary significantly depending on location and biomass productivity and that integration of novel dewatering equipment, order of operations, and equipment scaling can also have significant impacts on economics.

  13. Assessment of Algal Farm Designs Using a Dynamic Modular Approach

    SciTech Connect (OSTI)

    Abodeely, Jared; Coleman, Andre M.; Stevens, Daniel M.; Ray, Allison E.; Cafferty, Kara G.; Newby, Deborah T.

    2014-07-01

    The notion of renewable energy provides an important mechanism for diversifying an energy portfolio, which ultimately would have numerous benefits including increased energy resilience, reduction of foreign energy supplies, reduced GHG emissions, development of a green energy sector that contributes to economic growth, and providing a sustainable energy supply. The conversion of autotrophic algae to liquid transportation fuels is the basis of several decades of research to competitively bring energy-scale production into reality; however, many challenges still remain for making algal biofuels economically viable. Addressing current challenges associated with algal production systems, in part, requires the ability to assess spatial and temporal variability, rapidly evaluate alternative algal production system designs, and perform large-scale assessments considering multiple scenarios for thousands of potential sites. We introduce the Algae Logistics Model (ALM) which helps to address these challenges. The flexible nature of the ALM architecture allows the model to: 1) interface with external biomass production and resource assessment models, as well as other relevant datasets including those with spatiotemporal granularity; 2) interchange design processes to enable operational and economic assessments of multiple design configurations, including the integration of current and new innovative technologies; and 3) conduct trade-off analysis to help understand the site-specific techno-economic trade-offs and inform technology decisions. This study uses the ALM to investigate a baseline open-pond production system determined by model harmonization efforts conducted by the U.S. Department of Energy. Six sites in the U.S. southern-tier were sub-selected and assessed using daily site-specific algae biomass productivity data to determine the economic viability of large-scale open-pond systems. Results show that costs can vary significantly depending on location and biomass productivity and that integration of novel dewatering equipment, order of operations, and equipment scaling can also have significant impacts on economics.

  14. Magnitude and reactivity consequences of moisture ingress into the modular High-Temperature Gas-Cooled Reactor core

    SciTech Connect (OSTI)

    Smith, O.L. (Oak Ridge National Lab., TN (United States))

    1992-12-01

    Inadvertent admission of moisture into the primary system of a modular high-temperature gas-cooled reactor has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moisture-ingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. The rate and magnitude of steam buildup are found to be dominated by major system features such as break size compared with safety valve capacity and reliability and less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions.

  15. 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

  16. Liquefaction Reactor Design: April 5, 2013 Knorr, D.; Lukas,...

    Office of Scientific and Technical Information (OSTI)

    Production of Advanced Biofuels via Liquefaction - Hydrothermal Liquefaction Reactor Design: April 5, 2013 Knorr, D.; Lukas, J.; Schoen, P. 09 BIOMASS FUELS BIOFUELS CONVERSION;...

  17. Space Reactor Radiation Shield Design Summary, for Information

    SciTech Connect (OSTI)

    EC Pheil

    2006-02-17

    The purpose of this letter is to provide a summary of the Prometheus space reactor radiation shield design status at the time of program restructuring.

  18. Applicability of GALE-86 Codes to Integral Pressurized Water Reactor designs

    SciTech Connect (OSTI)

    Geelhood, Kenneth J.; Rishel, Jeremy P.

    2012-06-01

    This report describes work that Pacific Northwest National Laboratory is doing to assist the U.S. Nuclear Regulatory Commission (NRC) Office of New Reactors (NRO) staff in their reviews of applications for nuclear power plants using new reactor core designs. These designs include small integral PWRs (IRIS, mPower, and NuScale reactor designs), HTGRs, (pebble-bed and prismatic-block modular reactor designs) and SFRs (4S and PRISM reactor designs). Under this specific task, PNNL will assist the NRC staff in reviewing the current versions of the GALE codes and identify features and limitations that would need to be modified to accommodate the technical review of iPWR and mPower® license applications and recommend specific changes to the code, NUREG-0017, and associated NRC guidance. This contract is necessary to support the licensing of iPWRs with a near-term focus on the B&W mPower® reactor design. While the focus of this review is on the mPower® reactor design, the review of the code and the scope of recommended changes consider a revision of the GALE codes that would make them universally applicable for other types of integral PWR designs. The results of a detailed comparison between PWR and iPWR designs are reported here. Also included is an investigation of the GALE code and its basis and a determination as to the applicability of each of the bases to an iPWR design. The issues investigated come from a list provided by NRC staff, the results of comparing the PWR and iPWR designs, the parameters identified as having a large impact on the code outputs from a recent sensitivity study and the main bases identified in NUREG-0017. This report will provide a summary of the gaps in the GALE codes as they relate to iPWR designs and for each gap will propose what work could be performed to fill that gap and create a version of GALE that is applicable to integral PWR designs.

  19. Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components

    SciTech Connect (OSTI)

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

    2013-05-17

    This report identifies a number of requirements for prognostics health management of passive systems in AdvSMRs, documents technical gaps in establishing a prototypical prognostic methodology for this purpose, and describes a preliminary research plan for addressing these technical gaps. AdvSMRs span multiple concepts; therefore a technology- and design-neutral approach is taken, with the focus being on characteristics that are likely to be common to all or several AdvSMR concepts. An evaluation of available literature is used to identify proposed concepts for AdvSMRs along with likely operational characteristics. Available operating experience of advanced reactors is used in identifying passive components that may be subject to degradation, materials likely to be used for these components, and potential modes of degradation of these components. This information helps in assessing measurement needs for PHM systems, as well as defining functional requirements of PHM systems. An assessment of current state-of-the-art approaches to measurements, sensors and instrumentation, diagnostics and prognostics is also documented. This state-of-the-art evaluation, combined with the requirements, may be used to identify technical gaps and research needs in the development, evaluation, and deployment of PHM systems for AdvSMRs. A preliminary research plan to address high-priority research needs for the deployment of PHM systems to AdvSMRs is described, with the objective being the demonstration of prototypic prognostics technology for passive components in AdvSMRs. Greater efficiency in achieving this objective can be gained through judicious selection of materials and degradation modes that are relevant to proposed AdvSMR concepts, and for which significant knowledge already exists. These selections were made based on multiple constraints including the analysis performed in this document, ready access to laboratory-scale facilities for materials testing and measurement, and potential synergies with other national laboratory and university partners.

  20. Probabilistic accident analysis of the Pebble Bed modular Reactor for use with risk informed regulation

    E-Print Network [OSTI]

    Savkina, Marina D., 1973-

    2004-01-01

    One of the major challenges to the successful deployment of new nuclear plants in the United States is the regulatory process, which is largely based on water-reactor technology. While ongoing and expected efforts to license ...

  1. A 48-month extended fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

    SciTech Connect (OSTI)

    Erighin, M. A. [Babcock and Wilcox Company, 109 Ramsey Place, Lynchburg, VA 24502 (United States)

    2012-07-01

    The B and W mPower{sup TM} reactor is a small, rail-shippable pressurized water reactor (PWR) with an integral once-through steam generator and an electric power output of 150 MW, which is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height, but otherwise standard, PWR assemblies with the familiar 17 x 17 fuel rod array on a 21.5 cm inter-assembly pitch. The B and W mPower core design and cycle management plan, which were performed using the Studsvik core design code suite, follow the pattern of a typical nuclear reactor fuel cycle design and analysis performed by most nuclear fuel management organizations, such as fuel vendors and utilities. However, B and W is offering a core loading and cycle management plan for four years of continuous power operations without refueling and without the hurdles of chemical shim. (authors)

  2. 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

    2010-09-01

    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).

  3. Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup, and Oxygen Separation Equipment; Task 1: Cost Estimates of Small Modular Systems

    SciTech Connect (OSTI)

    Nexant Inc.

    2006-05-01

    This deliverable is the Final Report for Task 1, Cost Estimates of Small Modular Systems, as part of NREL Award ACO-5-44027, ''Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup and Oxygen Separation Equipment''. Subtask 1.1 looked into processes and technologies that have been commercially built at both large and small scales, with three technologies, Fluidized Catalytic Cracking (FCC) of refinery gas oil, Steam Methane Reforming (SMR) of Natural Gas, and Natural Gas Liquids (NGL) Expanders, chosen for further investigation. These technologies were chosen due to their applicability relative to other technologies being considered by NREL for future commercial applications, such as indirect gasification and fluidized bed tar cracking. Research in this subject is driven by an interest in the impact that scaling has on the cost and major process unit designs for commercial technologies. Conclusions from the evaluations performed could be applied to other technologies being considered for modular or skid-mounted applications.

  4. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    SciTech Connect (OSTI)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

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

    SciTech Connect (OSTI)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01

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

  6. Design of a Modularized "Smart" Faade System Objective: Comprehensively research faade-adapted systems and devices for

    E-Print Network [OSTI]

    Wolberg, George

    -adapted systems and devices for maintaining building comfort and energy-efficiency, and design and build of these systems means they are a step backward when it comes to energy- efficient building envelopes. In recentDesign of a Modularized "Smart" Façade System Objective: Comprehensively research façade

  7. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    SciTech Connect (OSTI)

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  8. Annular core for Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    SciTech Connect (OSTI)

    Turner, R.F.; Baxter, A.M.; Stansfield, O.M.; Vollman, R.E.

    1987-08-01

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40% greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93-m tall. Fuel elements contain TRISO-coated microspheres of 19.8% enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above.

  9. INEEL/EXT-01-01623 MODULAR PEBBLE-BED REACTOR PROJECT

    E-Print Network [OSTI]

    Particle Batch Failure Probabilities Using an Integral Formulation 14 2.1.2 Fission Gas Release, CO Production and Fission Product Chemistry 15 2.1.2.1 Fission Gas and CO Release Model 15 2.1.2.2 Fission-AC07-99ID13727 #12;ii #12;iii Table of Contents 1.0 Introduction 1 2.0 Gas Reactor Fuel Performance

  10. The design of a compact integral medium size PWR : the CIRIS

    E-Print Network [OSTI]

    Shirvan, Koroush

    2010-01-01

    The International Reactor Innovative and Secure (IRIS) is an advanced medium size, modular integral light water reactor design, rated currently at 1000 MWt. IRIS design has been under development by over 20 organizations ...

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

    E-Print Network [OSTI]

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

    2006-01-01

    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 ...

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

    E-Print Network [OSTI]

    Apte, Sourabh V.

    DESIGN OF A MICROCHANNEL BASED SOLAR RECEIVER/REACTOR FOR METHANE-STEAM REFORMING Drost, K. J designs that permit increase in the overall solar to chemical conversion efficiency are needed of fuels is needed. Solar re- ceivers based on a volumetric or cavity receiver design are common [8, 9, 10

  13. Subcontract Report: Modular Combined Heat & Power System for Utica College: Design Specification

    SciTech Connect (OSTI)

    Rouse, Greg

    2007-09-01

    Utica College, located in Utica New York, intends to install an on-site power/cogeneration facility. The energy facility is to be factory pre-assembled, or pre- assembled in modules, to the fullest extent possible, and ready to install and interconnect at the College with minimal time and engineering needs. External connections will be limited to fuel supply, electrical output, potable makeup water as required and cooling and heat recovery systems. The proposed facility will consist of 4 self-contained, modular Cummins 330kW engine generators with heat recovery systems and the only external connections will be fuel supply, electrical outputs and cooling and heat recovery systems. This project was eventually cancelled due to changing DOE budget priorities, but the project engineers produced this system design specification in hopes that it may be useful in future endeavors.

  14. Re-design and Evaluation of a Modular fNIRS-Probe for Employment in Neuroimaging Applications

    E-Print Network [OSTI]

    Daraio, Chiara

    Re-design and Evaluation of a Modular fNIRS-Probe for Employment in Neuroimaging Applications Zürich, January 2013 Project Type MSc/BSc-Thesis Goal We recently built miniaturized sensor modules featuring silicon photo-multipliers (SiPM) for very low light detection in functional near-infrared

  15. 646 IEEE TRANSACTIONS ON ROBOTICS AND AUTOMATION, VOL. 13, NO. 5, OCTOBER 1997 Design of Assembly Systems for Modular Products

    E-Print Network [OSTI]

    Kusiak, Andrew

    systems. Given a family of modular products, designing low cost assembly systems is an important problem to the challenge of agile manufacturing, companies are striving to provide a large variety of products at a low problem of the assembly system is formulated and solved by a tabu search based algorithm. Index Terms

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

    E-Print Network [OSTI]

    Myers, Richard Allen

    2011-01-01

    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

  17. Advancing Small Modular Reactors: How We're Supporting Next-Gen...

    Energy Savers [EERE]

    reducing both capital costs and construction times. SMR designs also have built-in passive safety systems that use the natural circulation of air, water and steam to maintain...

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

    SciTech Connect (OSTI)

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

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  19. 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.

    2004-10-06

    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.

  20. Preliminary Development of a Work Breakdown Structure (WBS) for Small Modular Reactors (SMRs)

    SciTech Connect (OSTI)

    Harrison, Thomas J.; Moses, Rebecca J.; Flanagan, George F.

    2014-10-01

    In summary, this preliminary WBS serves as an initial basis for the capital cost component of the economic analysis of SMRs. This preliminary WBS comes from the known WBS for existing, large nuclear power plants and develops the methodology for accounting for the anticipated differences between the current large plants and the projected SMR designs.

  1. PEBBLE-BED NUCLEAR REACTOR SYSTEM PHYSICS AND FUEL UTILIZATION 

    E-Print Network [OSTI]

    Kelly, Ryan 1989-

    2011-04-20

    The Generation IV Pebble Bed Modular Reactor (PMBR) design may be used for electricity production, co-generation applications (industrial heat, hydrogen production, desalination, etc.), and could potentially eliminate some high level nuclear wastes...

  2. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    SciTech Connect (OSTI)

    Ruger, C.J.; Higgins, J.C.

    1993-11-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970`s and early 1980`s raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants.

  3. Assessment of Materials Issues for Light-Water Small Modular Reactors

    SciTech Connect (OSTI)

    Sandusky, David; Lunceford, Wayne; Bruemmer, Stephen M.; Catalan, Michael A.

    2013-02-01

    The primary objective of this report is to evaluate materials degradation issue unique to the operational environments of LWSMR. Concerns for specific primary system components and materials are identified based on the review of design information shared by mPower and NuScale. Direct comparisons are made to materials issues recognized for advanced large PWRs and research activities are recommended as needed. The issues identified are intended to improve the capability of industry to evaluate the significance of any degradation that might occur during long-term LWSMR operation and by extension affect the importance of future supporting R&D.

  4. An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

    SciTech Connect (OSTI)

    Scarangella, M. J. [Babcock and Wilcox Company, 109 Ramsey Place, Lynchburg, VA 24502 (United States)

    2012-07-01

    The B and W mPower{sup TM} reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

  5. Reactor Engineering Design | netl.doe.gov

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration wouldMassR&D100 Winners * Impacts on GlobalRachel2RateCaseElementsOxide Fuel CellsReactor

  6. Basis for NGNP Reactor Design Down-Selection

    SciTech Connect (OSTI)

    L.E. Demick

    2011-11-01

    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.

  7. Basis for NGNP Reactor Design Down-Selection

    SciTech Connect (OSTI)

    L.E. Demick

    2010-08-01

    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. Design of slurry reactor for indirect liquefaction applications

    SciTech Connect (OSTI)

    Prakash, A.; Bendale, P.G.

    1991-01-01

    The objective of this project is to design and model a conceptual slurry reactor for two indirect liquefaction applications; (1) production of methanol and (2) production of hydrocarbon fuels via Fischer-Tropsch route. A slurry reactor is defined here as a three-phase bubble column reactor using a fine catalyst particle suspension in a high molecular weight liquid. The feed gas is introduced through spargers. It then bubbles through the column providing the agitation necessary for catalyst suspension and mass transfer. The reactor models for the two processes have been formulated using computer simulation. Process data, kinetic and thermodynamic data, heat and mass transfer data and hydrodynamic data have been used in the mathematical models to describe the slurry reactor for each of the two processes. Available data from process development units and demonstration units were used to test and validate the models. Commercial size slurry reactors for methanol and Fischer-Tropsch synthesis were sized using reactor models developed in this report.

  9. Design of slurry reactor for indirect liquefaction applications. Final report

    SciTech Connect (OSTI)

    Prakash, A.; Bendale, P.G.

    1991-12-31

    The objective of this project is to design and model a conceptual slurry reactor for two indirect liquefaction applications; (1) production of methanol and (2) production of hydrocarbon fuels via Fischer-Tropsch route. A slurry reactor is defined here as a three-phase bubble column reactor using a fine catalyst particle suspension in a high molecular weight liquid. The feed gas is introduced through spargers. It then bubbles through the column providing the agitation necessary for catalyst suspension and mass transfer. The reactor models for the two processes have been formulated using computer simulation. Process data, kinetic and thermodynamic data, heat and mass transfer data and hydrodynamic data have been used in the mathematical models to describe the slurry reactor for each of the two processes. Available data from process development units and demonstration units were used to test and validate the models. Commercial size slurry reactors for methanol and Fischer-Tropsch synthesis were sized using reactor models developed in this report.

  10. Design and fabrication of a modular multi-material 3D printer

    E-Print Network [OSTI]

    Lan, Justin (Justin T.)

    2013-01-01

    This thesis presents 3DP-0, a modular, multi-material 3D printer. Currently, 3D printers available on the market are typically expensive and difficult to develop. In addition, the simultaneous use of multiple materials in ...

  11. Small Modular Reactors - SRSCRO

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTechtail.Theory ofDidDevelopment Top Scientific ImpactTechnologies |SiteSitesmr Small

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

    E-Print Network [OSTI]

    Yarsky, Peter

    2005-01-01

    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 ...

  13. Constructal method to optimize solar thermochemical reactor design

    SciTech Connect (OSTI)

    Tescari, S.; Mazet, N.; Neveu, P.

    2010-09-15

    The objective of this study is the geometrical optimization of a thermochemical reactor, which works simultaneously as solar collector and reactor. The heat (concentrated solar radiation) is supplied on a small peripheral surface and has to be dispersed in the entire reactive volume in order to activate the reaction all over the material. A similarity between this study and the point to volume problem analyzed by the constructal approach (Bejan, 2000) is evident. This approach was successfully applied to several domains, for example for the coupled mass and conductive heat transfer (Azoumah et al., 2004). Focusing on solar reactors, this work aims to apply constructal analysis to coupled conductive and radiative heat transfer. As a first step, the chemical reaction is represented by a uniform heat sink inside the material. The objective is to optimize the reactor geometry in order to maximize its efficiency. By using some hypothesis, a simplified solution is found. A parametric study provides the influence of different technical and operating parameters on the maximal efficiency and on the optimal shape. Different reactor designs (filled cylinder, cavity and honeycomb reactors) are compared, in order to determine the most efficient structure according to the operating conditions. Finally, these results are compared with a CFD model in order to validate the assumptions. (author)

  14. 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

  15. Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 1, Final report

    SciTech Connect (OSTI)

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

    1992-03-01

    This report contains a review of design studies for inertial confinement reactors. The first of three volumes briefly discusses the following: Introduction; Key objectives, requirements, and assumptions; Systems modeling and trade studies; Prometheus-L reactor plant design overview; Prometheus-H reactor plant design overview; Key technical issues and R&D requirements; Comparison of IFE designs; and study conclusions.

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

    SciTech Connect (OSTI)

    B. Boer; A. M. Ougouag

    2010-05-01

    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.

  17. Fusion reactor design | Princeton Plasma Physics Lab

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformation Current HABFESOpportunities Nuclear Physics (NP) NP HomeSciencesreactor design

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

    E-Print Network [OSTI]

    Whitman, Joshua (Joshua J.)

    2007-01-01

    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 ...

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

    E-Print Network [OSTI]

    Whitman, Joshua (Joshua J.)

    2009-01-01

    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 ...

  20. Design and Fabrication of In-Reactor Experiment to Measure Tritium...

    Office of Environmental Management (EM)

    Design and Fabrication of In-Reactor Experiment to Measure Tritium Release and Speciation from LiAlO2 and LiAlO2Zr Cermets Design and Fabrication of In-Reactor Experiment to...

  1. Project Profile: Modular and Scalable Baseload Molten Salt Plant...

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

    Modular and Scalable Baseload Molten Salt Plant Conceptual Design and Feasibility Project Profile: Modular and Scalable Baseload Molten Salt Plant Conceptual Design and Feasibility...

  2. Assigning Seismic Design Category to Large Reactors: A Case Study of the ATR

    Office of Energy Efficiency and Renewable Energy (EERE)

    Assigning Seismic Design Category to Large Reactors: A Case Study of the ATR Stuart Jensen October 21, 2014

  3. Nuclear Design of the HOMER-15 Mars Surface Fission Reactor

    SciTech Connect (OSTI)

    Poston, David I.

    2002-07-01

    The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heat pipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed - which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heat pipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heat pipes. Fission energy is conducted from the fuel pins to the heat pipes, which then carry the heat to the Stirling engine. This paper describes conceptual design and nuclear performance the HOMER-15 reactor. (author)

  4. 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

  5. A New Architecture for Man: The Modular, Prefabricated Buildings of Ernest J. Kump, Jr.

    E-Print Network [OSTI]

    Stiles, Elaine

    2013-01-01

    Hilp on a modular plan, a prefabricated house design forPark houses were a system of nested prefabricated modular

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

    SciTech Connect (OSTI)

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

    2011-07-17

    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.

  7. A Multi-Modular Neutronically Coupled Power Generation System 

    E-Print Network [OSTI]

    Patel, Vishal

    2012-07-16

    EOL End of life HT-IMMTR High Temperature Integrated Multi-Modular Thermal Reactor HTGR High Temperature Gas Cooled Reactor LWR Light water reactor LOCA Loss of coolant accident LOFA Loss of ow accident MOL Middle of life MWth Megawatts Thermal... upwards of 45%. In comparison, LWR water rankine cycles have e ciencies of about 33%, meaning the ScCO2 brayton cycle could be much more economical than an LWR rankine cycle. 1.4 Design Objectives The system is intended for use either autonomously...

  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

    2014-06-01

    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. 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 simulations as a fundamental design tool in developing a new prototype high pressure organometallic chemical vapor deposition (HPOMCVD) reactor for use in thin film crystal growth. The advantages of such a reactor

  10. Design Construction and Operation of a Supercritical Carbon Dioxide (sCO2) Loop for Investigation of Dry Cooling and Natural Circulation Potential for Use in Advanced Small Modular Reactors Utilizing sCO2 Power Conversion Cycles.

    SciTech Connect (OSTI)

    Middleton, Bobby D.; Rodriguez, Salvador B.; Carlson, Matthew David

    2015-11-01

    This report outlines the work completed for a Laboratory Directed Research and Development project at Sandia National Laboratories from October 2012 through September 2015. An experimental supercritical carbon dioxide (sCO 2 ) loop was designed, built, and o perated. The experimental work demonstrated that sCO 2 can be uti lized as the working fluid in an air - cooled, natural circulation configuration to transfer heat from a source to the ultimate heat sink, which is the surrounding ambient environment in most ca ses. The loop was also operated in an induction - heated, water - cooled configuration that allows for measurements of physical parameters that are difficult to isolate in the air - cooled configuration. Analysis included the development of two computational flu id dynamics models. Future work is anticipated to answer questions that were not covered in this project.

  11. Modular Inverter for Advanced Control Applications In the fall of 2003, a team of graduate students was assembled to design and construct a

    E-Print Network [OSTI]

    Kimball, Jonathan W.

    students was assembled to design and construct a research-grade inverter. The goal was to have available report contains all of the important specifications and design details required to use and repair modular. This work was supported by the Grainger Center for Electric Machinery and Electromechanics and was advised

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

    E-Print Network [OSTI]

    Carpenter, David Michael

    2006-01-01

    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 ...

  13. High Temperature Gas Reactors Andrew C. Kadak, Ph.D.

    E-Print Network [OSTI]

    Heat Exchanger Design · Core Power Distribution Monitoring · Pebble Flow Experiments · Non Reactors #12;GT-MHR Module General Arrangement #12;GT-MHR Combines Meltdown-Proof Advanced Reactor and Gas MWt and 600 MWt Cores #12;GT-MHR Flow Schematic #12;Flow through Power Conversion Vessel #12;Modular

  14. 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))

    1992-07-01

    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.

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

    DOE Patents [OSTI]

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

    1984-08-30

    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.

  16. 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)

    1985-01-01

    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.

  17. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    SciTech Connect (OSTI)

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

    1980-09-01

    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.

  18. 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 ...

  19. Symmetric modular torsatron

    DOE Patents [OSTI]

    Rome, J.A.; Harris, J.H.

    1984-01-01

    A fusion reactor device is provided in which the magnetic fields for plasma confinement in a toroidal configuration is produced by a plurality of symmetrical modular coils arranged to form a symmetric modular torsatron referred to as a symmotron. Each of the identical modular coils is helically deformed and comprise one field period of the torsatron. Helical segments of each coil are connected by means of toroidally directed windbacks which may also provide part of the vertical field required for positioning the plasma. The stray fields of the windback segments may be compensated by toroidal coils. A variety of magnetic confinement flux surface configurations may be produced by proper modulation of the winding pitch of the helical segments of the coils, as in a conventional torsatron, winding the helix on a noncircular cross section and varying the poloidal and radial location of the windbacks and the compensating toroidal ring coils.

  20. Design Windows for a He Cooled Fusion Reactor* Dai-Kai Sze and Ahmed Hassanein

    E-Print Network [OSTI]

    Harilal, S. S.

    Design Windows for a He Cooled Fusion Reactor* Dai-Kai Sze and Ahmed Hassanein Argonne National Laboratory 9700 South Cass Avenue, Argonne, IL 60439 EQUATIONDERIVATION ABSTRACT A design window concept is developed for a He-cooled fusion reactor blanket and divertor design. This concept allows study

  1. NGNP Project Regulatory Gap Analysis for Modular HTGRs

    SciTech Connect (OSTI)

    Wayne Moe

    2011-09-01

    The Next Generation Nuclear Plant (NGNP) Project Regulatory Gap Analysis (RGA) for High Temperature Gas-Cooled Reactors (HTGR) was conducted to evaluate existing regulatory requirements and guidance against the design characteristics specific to a generic modular HTGR. This final report presents results and identifies regulatory gaps concerning current Nuclear Regulatory Commission (NRC) licensing requirements that apply to the modular HTGR design concept. This report contains appendices that highlight important HTGR licensing issues that were found during the RGA study. The information contained in this report will be used to further efforts in reconciling HTGR-related gaps in the NRC licensing structure, which has to date largely focused on light water reactor technology.

  2. Evaluation of Potential Locations for Siting Small Modular Reactors near Federal Energy Clusters to Support Federal Clean Energy Goals

    SciTech Connect (OSTI)

    Belles, Randy J.; Omitaomu, Olufemi A.

    2014-09-01

    Geographic information systems (GIS) technology was applied to analyze federal energy demand across the contiguous US. Several federal energy clusters were previously identified, including Hampton Roads, Virginia, which was subsequently studied in detail. This study provides an analysis of three additional diverse federal energy clusters. The analysis shows that there are potential sites in various federal energy clusters that could be evaluated further for placement of an integral pressurized-water reactor (iPWR) to support meeting federal clean energy goals.

  3. 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...

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

    SciTech Connect (OSTI)

    Pan, Paul Y

    2010-12-10

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

  5. One pass core design of a super fast reactor

    SciTech Connect (OSTI)

    Liu, Qingjie; Oka, Yoshiaki

    2013-07-01

    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)

  6. Core design studies for advanced burner test reactor.

    SciTech Connect (OSTI)

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

    2008-01-01

    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

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

    SciTech Connect (OSTI)

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

    1998-05-01

    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.

  8. Design, fabrication and testing of ModBot, the biomimetic, backdrivable, modular finger robot

    E-Print Network [OSTI]

    Kelley, Michael Scott

    2011-01-01

    of the design was the weight of the onboard gear head motorsuse gears. The second contribution is to optimize the design

  9. Initial Design of a Dual Fluidized Bed Reactor

    E-Print Network [OSTI]

    Yun, Minyoung

    2014-01-01

    the gasification processes in a BFB coal gasifier, chemicalby a partition wall in the BFB reactor. Solid and gas mixingand right compartment in the BFB reactor. A larger space is

  10. Advanced Test Reactor Design Basis Reconstitution Project Issue Resolution Process

    SciTech Connect (OSTI)

    Steven D. Winter; Gregg L. Sharp; William E. Kohn; Richard T. McCracken

    2007-05-01

    The Advanced Test Reactor (ATR) Design Basis Reconstitution Program (DBRP) is a structured assessment and reconstitution of the design basis for the ATR. The DBRP is designed to establish and document the ties between the Document Safety Analysis (DSA), design basis, and actual system configurations. Where the DBRP assessment team cannot establish a link between these three major elements, a gap is identified. Resolutions to identified gaps represent configuration management and design basis recovery actions. The proposed paper discusses the process being applied to define, evaluate, report, and address gaps that are identified through the ATR DBRP. Design basis verification may be performed or required for a nuclear facility safety basis on various levels. The process is applicable to large-scale design basis reconstitution efforts, such as the ATR DBRP, or may be scaled for application on smaller projects. The concepts are applicable to long-term maintenance of a nuclear facility safety basis and recovery of degraded safety basis components. The ATR DBRP assessment team has observed numerous examples where a clear and accurate link between the DSA, design basis, and actual system configuration was not immediately identifiable in supporting documentation. As a result, a systematic approach to effectively document, prioritize, and evaluate each observation is required. The DBRP issue resolution process provides direction for consistent identification, documentation, categorization, and evaluation, and where applicable, entry into the determination process for a potential inadequacy in the safety analysis (PISA). The issue resolution process is a key element for execution of the DBRP. Application of the process facilitates collection, assessment, and reporting of issues identified by the DBRP team. Application of the process results in an organized database of safety basis gaps and prioritized corrective action planning and resolution. The DBRP team follows the ATR DBRP issue resolution process which provides a method for the team to promptly sort and prioritize questions and issues between those that can be addressed as a normal part of the reconstitution project and those that are to be handle as PISAs. Presentation of the DBRP issue resolution process provides an example for similar activities that may be required at other facilities within the Department of Energy complex.

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

    DOE Patents [OSTI]

    Forsberg, Charles W. (Kingston, TN)

    1987-01-01

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

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

    DOE Patents [OSTI]

    Forsberg, C.W.

    1985-02-19

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

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

    E-Print Network [OSTI]

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

    2011-01-01

    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 ...

  14. 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)

    2014-09-30

    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.

  15. Design of a Modular Multilevel Converter as an Active Front-End for a magnet supply application

    E-Print Network [OSTI]

    Panagiotis, Asimakopoulos; Massimo, Bongiorno

    2015-01-01

    The aim of this work is to describe the general design procedure of a Modular Multilevel Converter (MMC) applied as an Active Front-End (AFE) for a magnet supply for beam accelerators. The dimensioning criteria for the converter and the dc-link capacitance are presented and the grid transformer requirements are set. Considering the converter design, the arm inductance calculation is based on the specifications for the arm-current ripple and the DC-link fault tolerance, but, also, on the limitation of the second harmonic and the second-order LC resonance of the arm current. The module capacitance value is evaluated by focusing on the required switching dynamics and the capacitor-voltage ripple according to a newly proposed graphical method. The loading of each semiconductor in the half bridge is calculated via simulation, indicating the unsymmetrical current distribution. It is concluded that the current distribution for each semiconductor depends on the mode of operation of the converter. The different criter...

  16. Modular optical detector system

    DOE Patents [OSTI]

    Horn, Brent A. (Livermore, CA); Renzi, Ronald F. (Tracy, CA)

    2006-02-14

    A modular optical detector system. The detector system is designed to detect the presence of molecules or molecular species by inducing fluorescence with exciting radiation and detecting the emitted fluorescence. Because the system is capable of accurately detecting and measuring picomolar concentrations it is ideally suited for use with microchemical analysis systems generally and capillary chromatographic systems in particular. By employing a modular design, the detector system provides both the ability to replace various elements of the detector system without requiring extensive realignment or recalibration of the components as well as minimal user interaction with the system. In addition, the modular concept provides for the use and addition of a wide variety of components, including optical elements (lenses and filters), light sources, and detection means, to fit particular needs.

  17. Modularity in Distributed Feature Composition Pamela Zave

    E-Print Network [OSTI]

    Greenberg, Albert

    , there is an abundance of experi- ence to draw upon. DFC was designed to support modular development of features; Section is an adaptation of the pipes-and-filters architectural style to telecommunication applications. This kind is an overview of pipes-and-filters modularity as realized in DFC. The benefit of feature modularity comes

  18. 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

    2004-01-01

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

  19. Design and prototyping of a modular human-powered swing carousel

    E-Print Network [OSTI]

    TenCate, Emily E

    2015-01-01

    The annual East Campus dormitory carnival at the Massachusetts Institute of Technology creates a unique learning experience by allowing undergraduate students to design and fabricate carnival rides for human riders. This ...

  20. 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)

    2014-09-30

    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)

  1. Systematic Design of Modular Products at Telemechanique Daniel E. Whitney, Liaison Scientist, Manufacturing

    E-Print Network [OSTI]

    Whitney, Daniel

    , Manufacturing Summary Telemechanique designs, builds, sells, and uses internally a wide variety of automation business of providing computer- integrated manufacturing systems. They are generally highly engineered as well as the focus of my visit. Because the products it sells are used in automation systems that make

  2. Designing a Component-Based Architecture for the Modeling and Simulation of Nuclear Fuels and Reactors

    E-Print Network [OSTI]

    Pennycook, Steve

    interest in nuclear energy in the U. S. Applications for 26 new reactors have been sub- mitted to the U. S. The NEAMS program is organized around four technical areas of the nuclear fuel cycle: fuels, reactorsDesigning a Component-Based Architecture for the Modeling and Simulation of Nuclear Fuels

  3. A BEHAVIOR-PRESERVING TRANSLATION FROM FBD DESIGN TO C IMPLEMENTATION FOR REACTOR

    E-Print Network [OSTI]

    of power plant construction technology. As the nuclear reactor protec- tion system (RPS) makes decisions- lator into C programs. KNICS (Korea Nuclear Instrumentation and Control System R&D Center) project [9A BEHAVIOR-PRESERVING TRANSLATION FROM FBD DESIGN TO C IMPLEMENTATION FOR REACTOR PROTECTION SYSTEM

  4. 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)

    2012-07-01

    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

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

    E-Print Network [OSTI]

    Forbes, Ian Alexander

    1970-01-01

    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 ...

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

    E-Print Network [OSTI]

    Bean, Malcolm K.

    2011-08-01

    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 ...

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

    E-Print Network [OSTI]

    Gibbs, Jonathan Paul

    2008-01-01

    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 ...

  8. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    SciTech Connect (OSTI)

    Gauld, Ian C; Sly, Nicholas C; Michel-Sendis, Franco

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

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

    SciTech Connect (OSTI)

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

    1992-03-01

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

  10. Comments on: Small Modular Reactors

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room News PublicationsAuditsCluster Compatibility Mode ClusterProteinReactions | Argonne

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

    E-Print Network [OSTI]

    Ellis, Tyler Shawn

    2009-01-01

    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 ...

  12. 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

    2008-04-23

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

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

    E-Print Network [OSTI]

    Petroski, Robert C

    2008-01-01

    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. Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup, and Oxygen Separation Equipment; Task 9: Mixed Alcohols From Syngas -- State of Technology

    SciTech Connect (OSTI)

    Nexant Inc.

    2006-05-01

    This deliverable is for Task 9, Mixed Alcohols from Syngas: State of Technology, as part of National Renewable Energy Laboratory (NREL) Award ACO-5-44027, ''Equipment Design and Cost Estimation for Small Modular Biomass Systems, Synthesis Gas Cleanup and Oxygen Separation Equipment''. Task 9 supplements the work previously done by NREL in the mixed alcohols section of the 2003 technical report Preliminary Screening--Technical and Economic Assessment of Synthesis Gas to Fuels and Chemicals with Emphasis on the Potential for Biomass-Derived Syngas.

  15. Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

    SciTech Connect (OSTI)

    Pope, Michael A.; Gougar, Hans D.; Ryskamp, J. M.

    2015-03-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

  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-

    2004-01-01

    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

    2010-01-01

    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. 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

    2010-09-01

    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.

  19. Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor

    SciTech Connect (OSTI)

    Ishii, M.; Xu, Y.; Revankar, S.T.

    2002-07-01

    A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified boiling water reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified boiling water reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident. (authors)

  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.

    1985-01-01

    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. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    SciTech Connect (OSTI)

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team indicates that the proposed solutions to the investigated operating cycle length barriers are both feasible and consistent with sound design practice.

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

    SciTech Connect (OSTI)

    John D. Bess; Margaret A. Marshall

    2013-02-01

    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. XAUV : modular high maneuverability autonomous underwater vehicle

    E-Print Network [OSTI]

    Walker, Daniel G. (Daniel George)

    2009-01-01

    The design and construction of a modular test bed autonomous underwater vehicle (AUV) is analyzed. Although a relatively common stacked-hull design is used, the state of the art is advanced through an aggressive power ...

  4. Design and performance of a high-pressure Fischer-Tropsch fluidized bed reactor

    SciTech Connect (OSTI)

    Weimer, A.W.; Quarderer, G.J.; Cochran, G.A.; Conway, M.M. )

    1988-01-01

    A 900 kg/day, CO/H/sub 2/, high-pressure, fluidized bed, pilot reactor was designed from first principles to achieve high reactant conversions and heat removal rates for the Fischer-Tropsch (F-T) synthesis of liquefied petroleum gases (LPG's). Suppressed bubble growth at high pressure allowed high reactant conversions which nearly matched those obtained at identical conditions in a lab scale fixed bed reactor. For GHSV approximately 1400 hr/sup -1/ and T = 658 {Kappa} at P approximately 7000 {kappa}Pa, reactant conversion exceeded 75%. The reactor heat removal capability exceeded twice design performance with the fluidized bed easily operating under thermally stable conditions. The fluidized catalyst was a potassium promoted, molybdenum on carbon (Mo/{Kappa}/C) catalyst which did not produce any detrimental waxy products. Long catalyst lifetimes of 1000 hrs on steam between regenerations allowed the fluidized bed to be operated in a batch mode.

  5. Update; Sodium advanced fast reactor (SAFR) concept

    SciTech Connect (OSTI)

    Oldenkamp, R.D.; Brunings, J.E. ); Guenther, E. ); Hren, R. )

    1988-01-01

    This paper reports on the sodium advanced fast reactor (SAFR) concept developed by the team of Rockwell International, Combustion Engineering, and Bechtel during the 3-year period extending from January 1985 to December 1987 as one element in the U.S. Department of Energy's Advanced Liquid Metal Reactor Program. In January 1988, the team was expanded to include Duke Engineering and Services, Inc., and the concept development was extended under DOE's Program for Improvement in Advanced Modular LMR Design. The SAFR plant concept employs a 450-MWe pool-type liquid metal cooled reactor as its basic module. The reactor assembly module is a standardized shop-fabricated unit that can be shipped to the plant site by barge for installation. Shop fabrication minimizes nuclear-grade field fabrication and reduces the plant construction schedule. Reactor modules can be used individually or in multiples at a given site to supply the needed generating capacity.

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

    DOE Patents [OSTI]

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

    1993-01-01

    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.

  7. Upflow column reactor design for dechlorination of chlorinated pulping wastes by Penicillium camemberti

    E-Print Network [OSTI]

    Taseli, Hasan

    . Introduction The main constraints of effluent discharges from the pulp and paper mill industry are adsorbable technology that permits continuous and efficient treatment of chlorinated effluents from the paper industryUpflow column reactor design for dechlorination of chlorinated pulping wastes by Penicillium

  8. Research and Development Assessments for Prometheus eavy Ion and Laser Driven Inertial Fusion Energy Reactor Designs

    E-Print Network [OSTI]

    Abdou, Mohamed

    -ion and laser drivers. INTRODUCTION Two commercial central station electric power plants have been conceptually designed and analyzed in the Prometheus[11study led by McDonnell Douglas Aerospace. These plants use reactors, (2) provide programmatic-decision makers with a list of important R&D tasks that need

  9. 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)

    2006-01-20

    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.

  10. Modular radiochemistry synthesis system

    DOE Patents [OSTI]

    Satyamurthy, Nagichettiar; Barrio, Jorge R.; Amarasekera, Bernard; Van Dam, R. Michael; Olma, Sebastian; Williams, Dirk; Eddings, Mark; Shen, Clifton Kwang-Fu

    2015-12-15

    A modular chemical production system includes multiple modules for performing a chemical reaction, particularly of radiochemical compounds, from a remote location. One embodiment comprises a reaction vessel including a moveable heat source with the position thereof relative to the reaction vessel being controllable from a remote position. Alternatively the heat source may be fixed in location and the reaction vial is moveable into and out of the heat source. The reaction vessel has one or more sealing plugs, the positioning of which in relationship to the reaction vessel is controllable from a remote position. Also the one or more reaction vessel sealing plugs can include one or more conduits there through for delivery of reactants, gases at atmospheric or an elevated pressure, inert gases, drawing a vacuum and removal of reaction end products to and from the reaction vial, the reaction vial with sealing plug in position being operable at elevated pressures. The modular chemical production system is assembled from modules which can each include operating condition sensors and controllers configured for monitoring and controlling the individual modules and the assembled system from a remote position. Other modules include, but are not limited to a Reagent Storage and Delivery Module, a Cartridge Purification Module, a Microwave Reaction Module, an External QC/Analysis/Purification Interface Module, an Aliquotting Module, an F-18 Drying Module, a Concentration Module, a Radiation Counting Module, and a Capillary Reactor Module.

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

    SciTech Connect (OSTI)

    Jon Carmack; Kemal Pasamehmetoglu

    2010-07-01

    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.

  12. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    SciTech Connect (OSTI)

    Pope, R B; Diggs, J M [eds.

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

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

    SciTech Connect (OSTI)

    Sabharwall, P.; Kim, E. S.; Siahpush, A.; McKellar, M.; Patterson, M.

    2012-07-01

    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)

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

    SciTech Connect (OSTI)

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

    2011-07-17

    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.

  15. 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

    2009-05-15

    Approved by: Chair of Committee, Kenneth L. Peddicord Committee Members, Marvin L. Adams William H. Marlow Karl T. Hartwig Edwin A. Harvego Head of Department, William E. Burchill May 2007 Major Subject: Nuclear... with the Idaho National Laboratory and General Atomics for my research. It was a great learning experience for me to work on this project with Dr. Peddicord and under the arrangements he made for me. I would like to thank Marvin L. Adams, William H. Marlow...

  16. Preliminary design of a fusion reactor fuel cleanup system by the palladium-alloy membrane method

    SciTech Connect (OSTI)

    Yoshida, H.; Konishi, S.; Naruse, Y.

    1983-05-01

    A design for a palladium diffuser and fuel cleanup system for a deuterium-tritium fusion reactor is proposed. The feasibility of the palladium-alloy membrane method is discussed based on early studies by the authors. Operating conditions of the palladium diffuser are determined experimentally. Dimensions of the diffuser are estimated from computer simulation. A fuel cleanup system is designed under the feed conditions of the Tritium Systems Test Assembly at Los Alamos National Laboratory. The system is composed of palladium diffusers, catalytic oxidizer, freezer, and zinc beds and has some advantages in system layout and operation. This design can readily be extended to other conditions of plasma exhaust gases.

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

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01

    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.

  18. 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.

    1980-07-01

    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.

  19. Modular robot

    DOE Patents [OSTI]

    Ferrante, T.A.

    1997-11-11

    A modular robot may comprise a main body having a structure defined by a plurality of stackable modules. The stackable modules may comprise a manifold, a valve module, and a control module. The manifold may comprise a top surface and a bottom surface having a plurality of fluid passages contained therein, at least one of the plurality of fluid passages terminating in a valve port located on the bottom surface of the manifold. The valve module is removably connected to the manifold and selectively fluidically connects the plurality of fluid passages contained in the manifold to a supply of pressurized fluid and to a vent. The control module is removably connected to the valve module and actuates the valve module to selectively control a flow of pressurized fluid through different ones of the plurality of fluid passages in the manifold. The manifold, valve module, and control module are mounted together in a sandwich-like manner and comprise a main body. A plurality of leg assemblies are removably connected to the main body and are removably fluidically connected to the fluid passages in the manifold so that each of the leg assemblies can be selectively actuated by the flow of pressurized fluid in different ones of the plurality of fluid passages in the manifold. 12 figs.

  20. Modular robot

    DOE Patents [OSTI]

    Ferrante, Todd A. (Idaho Falls, ID)

    1997-01-01

    A modular robot may comprise a main body having a structure defined by a plurality of stackable modules. The stackable modules may comprise a manifold, a valve module, and a control module. The manifold may comprise a top surface and a bottom surface having a plurality of fluid passages contained therein, at least one of the plurality of fluid passages terminating in a valve port located on the bottom surface of the manifold. The valve module is removably connected to the manifold and selectively fluidically connects the plurality of fluid passages contained in the manifold to a supply of pressurized fluid and to a vent. The control module is removably connected to the valve module and actuates the valve module to selectively control a flow of pressurized fluid through different ones of the plurality of fluid passages in the manifold. The manifold, valve module, and control module are mounted together in a sandwich-like manner and comprise a main body. A plurality of leg assemblies are removably connected to the main body and are removably fluidically connected to the fluid passages in the manifold so that each of the leg assemblies can be selectively actuated by the flow of pressurized fluid in different ones of the plurality of fluid passages in the manifold.

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    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,

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

    E-Print Network [OSTI]

    Myers, Richard Allen

    2011-01-01

    exceed that of fusion power reactors for wall loadings up toplasma. The vessel wall temperatures in power reactors isfor· the first wall of fusion power reactors. However, for

  3. 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.

    2012-12-01

    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.

  4. Safety design approach for external events in Japan sodium-cooled fast reactor

    SciTech Connect (OSTI)

    Yamano, H.; Kubo, S. [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki, 311-1393 (Japan); Tani, A. [Mitsubishi FBR Systems, Inc., 2-34-17, Jingumae, Shibuya-ku, Tokyo, 150-0001 (Japan); Nishino, H.; Sakai, T. [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki, 311-1393 (Japan)

    2012-07-01

    This paper describes a safety design approach for external events in the design study of Japan sodium-cooled fast reactor. An emphasis is introduction of a design extension external condition (DEEC). In addition to seismic design, other external events such as tsunami, strong wind, abnormal temperature, etc. were addressed in this study. From a wide variety of external events consisting of natural hazards and human-induced ones, a screening method was developed in terms of siting, consequence, frequency to select representative events. Design approaches for these events were categorized on the probabilistic, statistical and deterministic basis. External hazard conditions were considered mainly for DEECs. In the probabilistic approach, the DEECs of earthquake, tsunami and strong wind were defined as 1/10 of exceedance probability of the external design bases. The other representative DEECs were also defined based on statistical or deterministic approaches. (authors)

  5. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    SciTech Connect (OSTI)

    Lv, Quiping; Sun, Xiaodong; Chtistensen, Richard; Blue, Thomas; Yoder, Graydon; Wilson, Dane

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  6. Role of Nuclear Grade Graphite in Oxidation in Modular HTGRs

    SciTech Connect (OSTI)

    Willaim Windes; G. Strydom; J. Kane; R. Smith

    2014-11-01

    The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of core environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.

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

    SciTech Connect (OSTI)

    Frank Darmann; Robert Lombaerde; Franco Moriconi; Albert Nelson

    2011-10-31

    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? 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? 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? 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.

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    in particular, in boiling water reactors) there is a strongdirect contact boiling water fast reactor (PBWFR)”. In:

  9. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

    SciTech Connect (OSTI)

    George, J.A.

    1991-09-01

    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

  10. 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.)

    2014-01-01

    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 ...

  11. Portable modular detection system

    DOE Patents [OSTI]

    Brennan, James S. (Rodeo, CA); Singh, Anup (Danville, CA); Throckmorton, Daniel J. (Tracy, CA); Stamps, James F. (Livermore, CA)

    2009-10-13

    Disclosed herein are portable and modular detection devices and systems for detecting electromagnetic radiation, such as fluorescence, from an analyte which comprises at least one optical element removably attached to at least one alignment rail. Also disclosed are modular detection devices and systems having an integrated lock-in amplifier and spatial filter and assay methods using the portable and modular detection devices.

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

    SciTech Connect (OSTI)

    Michael G. McKellar; Edwin A. Harvego

    2010-05-01

    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.

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

    SciTech Connect (OSTI)

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

    2004-10-01

    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.

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

    SciTech Connect (OSTI)

    Adams, S.R.

    1985-10-01

    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.

  15. A Compact Quasi-axisymmetric Stellarator Reactor

    SciTech Connect (OSTI)

    L.P. Ku; the ARIES-CS Team

    2003-10-20

    We report the progress made in assessing the potential of compact, quasi-axisymmetric stellarators as power-producing reactors. Using an aspect ratio A=4.5 configuration derived from NCSX and optimized with respect to the quasi-axisymmetry and MHD stability in the linear regime as an example, we show that a reactor of 1 GW(e) maybe realizable with a major radius *8 m. This is significantly smaller than the designs of stellarator reactors attempted before. We further show the design of modular coils and discuss the optimization of coil aspect ratios in order to accommodate the blanket for tritium breeding and radiation shielding for coil protection. In addition, we discuss the effects of coil aspect ratio on the peak magnetic field in the coils.

  16. Preliminary design of ultra-long cycle fast reactor employing breed-and-burn strategy

    SciTech Connect (OSTI)

    Tak, T. W.; Yu, H.; Kim, J. H.; Lee, D. [Ulsan National Inst. of Science and Technology, gil 50, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of); Kim, T. K. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60564 (United States)

    2012-07-01

    A new design of ultra-long cycle fast reactor with power rate of 1000 MWe (UCFR) has been developed based on the strategy of breed-and burn. The bottom region of the core with low enriched uranium (LEU) plays a role of igniter of the core burning and the upper natural uranium (NU) region acts as blanket for breeding. Fissile materials are bred in the blanket and the active core moves upward at a speed of 5.4 cm/year. Through the core depletion calculation using Monte Carlo code, McCARD, it is confirmed that a full power operation of 60 years without refueling is feasible. Core performance characteristics have been evaluated in terms of axial/radial power shapes, reactivity feedback coefficients, etc. This design will serve as a base model for further design study of UCFRs using LWR spent fuels in the blanket region. (authors)

  17. 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)

    2012-07-01

    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)

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    increase for a typical sodium fast reactor fuel rod geometryof the new Russian sodium fast reactor BN-800 [111]. Thethe strong focus of sodium fast reactor research to avoid

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    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

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    reactor pas- sive reactivity control systems: LEM and LIM”. In: Nuclearnuclear reactor. Also it did not feature oxygen content controlControl Rods Reactor Concept and Plant Dynamics Analyses”. In: Journal of Nuclear

  1. Design of a new reactor-like high temperature near ambient pressure scanning tunneling microscope for catalysis studies

    E-Print Network [OSTI]

    Tao, Franklin Feng; Nguyen, Luan; Zhang, Shiran

    2013-01-01

    Here, we present the design of a new reactor-like high-temperature near ambient pressure scanning tunneling microscope (HT-NAP-STM) for catalysis studies. This HT-NAP-STM was designed for exploration of structures of catalyst surfaces at atomic...

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

    SciTech Connect (OSTI)

    Bruce G. Schnitzler

    2012-01-01

    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.

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

    SciTech Connect (OSTI)

    Ingesson, L. C.; Collaboration: F4E Diagnostic Project Team

    2014-08-21

    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.

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

    SciTech Connect (OSTI)

    Baker, C.C.

    1988-08-01

    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.

  5. Balance of Plant System Analysis and Component Design of Turbo...

    Office of Scientific and Technical Information (OSTI)

    Temperature Gas Reactor Systems The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve...

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

    SciTech Connect (OSTI)

    Bruce G. Schnitzler; Stanley K. Borowski

    2012-07-01

    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.

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

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01

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

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

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01

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

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

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01

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

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

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

    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.

  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

    2003-09-01

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

    SciTech Connect (OSTI)

    Philip Casey Durst; Mark Schanfein

    2012-08-01

    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.

  13. Modeling of the Reactor Core Isolation Cooling Reposnse to Beyond Design Basis Operations - Interim Report

    SciTech Connect (OSTI)

    Ross, Kyle; Cardoni, Jeffrey N.; Wilson, Chisom Shawn; Morrow, Charles; Osborn, Douglas; Gauntt, Randall O.

    2015-12-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine decreases the developed turbine torque; the RCIC speed then slows, and thus the pump flow rate to the RPV decreases. Subsequently, RPV water level decreases due to continued boiling and the liquid fraction flowing to the RCIC decreases, thereby accelerating the RCIC and refilling the RPV. The feedback cycle then repeats itself and/or reaches a quasi-steady equilibrium condition. In other words, the water carry-over is limited by cyclic RCIC performance degradation, and hence the system becomes self-regulating. The indications achieved to date with the system model are more qualitative than quantitative. The avenues being pursued to increase the fidelity of the model are expected to add quantitative realism. The end product will be generic in the sense that the RCIC model will be incorporable within the larger reactor coolant system model of any nuclear power plant or experimental configuration. ACKNOWLEDGEMENTS This work is funded through the U.S. Department of Energy - Office of Nuclear Energy's Light Water Reactor Sustainability Program. Clinton Smith of Phoenix Analysis & Design Technologies (PADT) is acknowledged for providing assistance on the FLUENT efforts in this report, as well as modifying the geometry model of the Terry turbine to make it more amenable for FLUENT analysis with rotating reference frames.

  14. Burnup concept for a long-life fast reactor core using MCNPX.

    SciTech Connect (OSTI)

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. 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. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  15. Initial Requirements for Gas-Cooled Fast Reactor (GFR) System Design, Performance, and Safety Analysis Models

    SciTech Connect (OSTI)

    Kevan D. Weaver; Thomas Y. C. Wei

    2004-08-01

    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 radio toxicity 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. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  16. 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

    2008-05-05

    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.

  17. A New Architecture for Man: The Modular, Prefabricated Buildings of Ernest J. Kump, Jr.

    E-Print Network [OSTI]

    Stiles, Elaine

    2013-01-01

    on a modular plan, a prefabricated house design for theirPark houses were a system of nested prefabricated modularknown prefabricated design was the 1945 “Prebuilt House” in

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

    SciTech Connect (OSTI)

    1997-04-18

    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.

  19. Design criteria of a chemical reactor based on a chaotic flow

    E-Print Network [OSTI]

    X. Z. Tang; A. H. Boozer

    1998-03-11

    We consider the design criteria of a chemical mixing device based on a chaotic flow, with an emphasis on the steady-state devices. The merit of a reactor, defined as the $Q$-factor, is related to the physical dimension of the device and the molecular diffusivity of the reactants through the local Lyapunov exponents of the flow. The local Lyapunov exponent can be calculated for any given flow field and it can also be measured in experimental situations. Easy-to-compute formulae are provided to estimate the $Q$-factor given either the exact spatial dependence of the local Lyapunov exponent or its probability distribution function. The requirements for optimization are made precise in the context of local Lyapunov exponents.

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

    E-Print Network [OSTI]

    Kempf, Stephanie Anne

    2011-01-01

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    high temperature reactors (VHTR), 3) supercritical waterup to 760 ° C. The DOE VHTR program is also sponsoring a

  2. SEAB Subcommittee on Small Modular Reactors (SMR)

    Energy Savers [EERE]

    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 on DeliciousMathematicsEnergyInterestedReplacement-2-A WholesaleRetrofitElectricalDepartment2-E WholesaleSEAB

  3. Small Modular Nuclear Reactors | Department of Energy

    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 on Delicious Rank EERE: Alternative FuelsofProgram: Report AppendicesAVideo »ServicesShale GasPrograms:Deployment

  4. Modular tokamak magnetic system

    DOE Patents [OSTI]

    Yang, Tien-Fang (Wayland, MA)

    1988-01-01

    A modular tokamak system comprised of a plurality of interlocking moldules. Each module is comprised of a vacuum vessel section, a toroidal field coil, moldular saddle coils which generate a poloidal magnetic field and ohmic heating coils.

  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

    2011-10-01

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

    SciTech Connect (OSTI)

    Pearlman, Howard; Chen, Chien-Hua

    2014-08-27

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

  7. Laminar Entrained Flow Reactor (Fact Sheet)

    SciTech Connect (OSTI)

    Not Available

    2014-02-01

    The Laminar Entrained Flow Reactor (LEFR) is a modular, lab scale, single-user reactor for the study of catalytic fast pyrolysis (CFP). This system can be employed to study a variety of reactor conditions for both in situ and ex situ CFP.

  8. Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel

    SciTech Connect (OSTI)

    Rearden, B.T. [Oak Ridge National Laboratory (United States); Anderson, W.J. [Framatome ANP, Inc. (France); Harms, G.A. [Sandia National Laboratories (United States)

    2005-08-15

    Framatome ANP, Sandia National Laboratories (SNL), Oak Ridge National Laboratory (ORNL), and the University of Florida are cooperating on the U.S. Department of Energy Nuclear Energy Research Initiative (NERI) project 2001-0124 to design, assemble, execute, analyze, and document a series of critical experiments to validate reactor physics and criticality safety codes for the analysis of commercial power reactor fuels consisting of UO{sub 2} with {sup 235}U enrichments {>=}5 wt%. The experiments will be conducted at the SNL Pulsed Reactor Facility.Framatome ANP and SNL produced two series of conceptual experiment designs based on typical parameters, such as fuel-to-moderator ratios, that meet the programmatic requirements of this project within the given restraints on available materials and facilities. ORNL used the Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) to assess, from a detailed physics-based perspective, the similarity of the experiment designs to the commercial systems they are intended to validate. Based on the results of the TSUNAMI analysis, one series of experiments was found to be preferable to the other and will provide significant new data for the validation of reactor physics and criticality safety codes.

  9. 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.

    2004-10-06

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1994-07-01

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1994-07-01

    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.

  12. Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2009-11-01

    An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, 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 high-enriched uranium fuel core. These studies 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.

  13. Multi-Applications Small Light Water Reactor - NERI Final Report

    SciTech Connect (OSTI)

    S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

    2003-12-01

    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.

  14. A U. S. Perspective on Fast Reactor Fuel Fabrication Technology and Experience Part I: Metal Fuels and Assembly Design

    SciTech Connect (OSTI)

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter; Douglas C. Crawford; Mitchell K. Meyer

    2009-06-01

    This paper is Part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II and the Fast Flux Test Facility, and it also refers to the impact of development in other nations. Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated into a foundation of research and resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    structure. The reactor uses carbide fuel, a com- posite ofvariables: - Fuel type (Oxide, Carbide, Nitride, Metallic) -99% N-15) Carbide Absorption in non-actinide fuel components

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    waste generation, uranium mining needs and proliferationis little need for uranium mining, since these reactors willis no further need for uranium mining. B&B technology thus

  17. 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

  18. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor

    SciTech Connect (OSTI)

    Leipold, F.; Furtula, V.; Salewski, M.; Bindslev, H.; Korsholm, S. B.; Meo, F.; Michelsen, P. K.; Moseev, D.; Nielsen, S. K.; Stejner, M. [Association Euratom--Risoe National Laboratory for Sustainable Energy, Technical University of Denmark, DK-4000 Roskilde (Denmark)

    2009-09-15

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic using gyrotrons operated at 60 GHz will meet the requirements for spatially and temporally resolved measurements of the velocity distributions of confined fast alphas in ITER by evaluating the scattered radiation (CTS signal). While a receiver antenna on the low field side of the tokamak, resolving near perpendicular (to the magnetic field) velocity components, has been enabled, an additional antenna on the high field side (HFS) would enable measurements of near parallel (to the magnetic field) velocity components. A compact design solution for the proposed mirror system on the HFS is presented. The HFS CTS antenna is located behind the blankets and views the plasma through the gap between two blanket modules. The viewing gap has been modified to dimensions 30x500 mm{sup 2} to optimize the CTS signal. A 1:1 mock-up of the HFS mirror system was built. Measurements of the beam characteristics for millimeter-waves at 60 GHz used in the mock-up agree well with the modeling.

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

    SciTech Connect (OSTI)

    Trianti, Nuri E-mail: szaki@fi.itba.c.id; Su'ud, Zaki E-mail: szaki@fi.itba.c.id; Arif, Idam E-mail: szaki@fi.itba.c.id; Riyana, EkaSapta

    2014-09-30

    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.

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

    E-Print Network [OSTI]

    Rubloff, Gary W.

    , University of Maryland, College Park, MD 20742 Gary W. Rubloff, Laurent Henn-Lecordier, and Yijun Liu Dept. of Materials Science and Engineering and Institute for Systems Research, University of Maryland, College Park with the semiconductor industry, from early bell-jar CVD reactors to current cold-wall single-wafer reactors (Xia et al

  1. Human Factors Aspects of Operating Small Reactors

    SciTech Connect (OSTI)

    OHara, J.M.; Higgins, J.; Deem, R.; Xing, J.; DAgostino, A.

    2010-11-07

    The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. They are considering small modular reactors (SMRs) as one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants, and so may require a concept of operations (ConOps) that also is different. The U.S. Nuclear Regulatory Commission (NRC) has begun examining the human factors engineering- (HFE) and ConOps- aspects of SMRs; if needed, they will formulate guidance to support SMR licensing reviews. We developed a ConOps model, consisting of the following dimensions: Plant mission; roles and responsibilities of all agents; staffing, qualifications, and training; management of normal operations; management of off-normal conditions and emergencies; and, management of maintenance and modifications. We are reviewing information on SMR design to obtain data about each of these dimensions, and have identified several preliminary issues. In addition, we are obtaining operations-related information from other types of multi-module systems, such as refineries, to identify lessons learned from their experience. Here, we describe the project's methodology and our preliminary findings.

  2. 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

    2012-09-01

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

  3. Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins

    SciTech Connect (OSTI)

    Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara; Peters, Curtis [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2005-02-06

    Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an early prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore called 'HPR-1'.

  4. 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.; Wemple, C.A.; Jahshan, S.N.; Ryskammp, J.M.

    1995-08-01

    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.

  5. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    SciTech Connect (OSTI)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  6. DOI Designates B Reactor at DOE's Hanford Site as a National...

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

    plutonium for the atomic weapon that was dropped on Nagasaki, Japan to end World War II (WWII). "B Reactor has a special feeling and association - as a landmark should. For...

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

    E-Print Network [OSTI]

    Nguyen, Nghia T. (Nghia Tat)

    2014-01-01

    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 ...

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

    E-Print Network [OSTI]

    Xu, Zhiwen, 1975-

    2003-01-01

    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 ...

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

    E-Print Network [OSTI]

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

    2010-01-01

    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 ...

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

    E-Print Network [OSTI]

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

    2007-01-01

    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 ...

  11. 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

    2008-01-01

    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 ...

  12. Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor

    Energy Savers [EERE]

    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 on QA:QA J-E-1 SECTION J APPENDIX E LIST OF APPLICABLE DIRECTIVES PursuantEnergy Small Column Ion ExchangeofofVessel

  13. Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor

    Energy Savers [EERE]

    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 on QA:QA J-E-1 SECTION J APPENDIX E LIST OF APPLICABLE DIRECTIVES PursuantEnergy Small Column Ion

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

    SciTech Connect (OSTI)

    Yoder, G.L.

    2005-10-03

    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.

  15. Modular Coil Sys Requirements

    E-Print Network [OSTI]

    Princeton Plasma Physics Laboratory

    Modular Coil Sys Requirements BSPEC140101 Conductor Specification CSPEC1420301 Winding Form Specification CSPEC1410313 Mod Coil Asm Specification CSPEC1420501 Coil TypeAB Winding Pack SE142A019R0 Coil Asm SE141058R3 Pol Break Shim Asm SE141078R2A EM/Struct Analysis of Coil and Shell CALC1400100 Mod

  16. A vectorized heat transfer model for solid reactor cores

    SciTech Connect (OSTI)

    Rider, W.J.; Cappiello, M.W.; Liles, D.R.

    1990-01-01

    The new generation of nuclear reactors includes designs that are significantly different from light water reactors. Among these new reactor designs is the Modular High-Temperature Gas-Cooled Reactor (MHTGR). In addition, nuclear thermal rockets share a number of similarities with terrestrial HTGRs and would be amenable to similar types of analyses. In these reactors, the heat transfer in the solid core mass is of primary interest in design and safety assessment. One significant safety feature of these reactors is the capability to withstand a loss of pressure and forced cooling in the primary system and still maintain peak fuel temperatures below the safe threshold for retaining the fission products. To accurately assess the performance of gas-cooled reactors during these types of transients, a Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions. Also, HERA has been used to analyze a depressurized loss of forced cooling transient in a HTGR with a very detailed three-dimensional input model. The results compare favorably with other means of analysis and provide further validation of the models and methods. 18 refs., 11 figs.

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

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01

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

  18. 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

    2011-02-01

    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.

  19. Steps towards verification and validation of the Fetch code for Level 2 analysis, design, and optimization of aqueous homogeneous reactors

    SciTech Connect (OSTI)

    Nygaard, E. T.; Pain, C. C.; Eaton, M. D.; Gomes, J. L. M. A.; Goddard, A. J. H.; Gorman, G.; Tollit, B.; Buchan, A. G.; Cooling, C. M.; Angelo, P. L.

    2012-07-01

    Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' While AHRs have been modeled effectively using simplified 'Level 1' tools, the complex interactions between fluids, neutronics, and solid structures are important (but not necessarily safety significant). These interactions require a 'Level 2' modeling tool. Imperial College London (ICL) has developed such a tool: Finite Element Transient Criticality (FETCH). FETCH couples the radiation transport code EVENT with the computational fluid dynamics code (Fluidity), the result is a code capable of modeling sub-critical, critical, and super-critical solutions in both two-and three-dimensions. Using FETCH, ICL researchers and B and W engineers have studied many fissioning solution systems include the Tokaimura criticality accident, the Y12 accident, SILENE, TRACY, and SUPO. These modeling efforts will ultimately be incorporated into FETCH'S extensive automated verification and validation (V and V) test suite expanding FETCH'S area of applicability to include all relevant physics associated with AHRs. These efforts parallel B and W's engineering effort to design and optimize an AHR to produce Mo99. (authors)

  20. Design change management in regulation of nuclear fleets: World nuclear association's working groups on Cooperation in Reactor Design Evaluation and Licensing (CORDEL)

    SciTech Connect (OSTI)

    Swinburn, R. [CORDEL DCM Task Force, Rolls-Royce Plc (United Kingdom); Borysova, I. [CORDEL, WNA, 22a St.James Sq., London SW1Y 4JH (United Kingdom); Waddington, J. [CORDEL Group (United Kingdom); Head, J. G. [CORDEL Group, GE-Hitachi Nuclear Energy (United Kingdom); Raidis, Z. [CORDEL Group, Candu Energy (United Kingdom)

    2012-07-01

    The 60 year life of a reactor means that a plant will undergo change during its life. To ensure continuing safety, changes must be made with a full understanding of the design intent. With this aim, regulators require that each operating organisation should have a formally designated entity responsible for complete design knowledge in regard to plant safety. INSAG-19 calls such an entity 'Design Authority'. This requirement is difficult to achieve, especially as the number of countries and utilities operating plants increases. Some of these operating organisations will be new, and some will be small. For Gen III plants sold on a turnkey basis, it is even more challenging for the operating company to develop and retain the full knowledge needed for this role. CORDEL's Task Force entitled 'Design Change Management' is investigating options for effective design change management with the aim to support design standardization throughout a fleet's lifetime by means of enhanced international cooperation within industry and regulators. This paper starts with considering the causes of design change and identifies reasons for the increased beneficial involvement of the plant's original vendor in the design change process. A key central theme running through the paper is the definition of responsibilities for design change. Various existing mechanisms of vendor-operator interfaces over design change and how they are managed in different organisational and regulatory environments around the world are considered, with the functionality of Owners Groups and Design Authority being central. The roles played in the design change process by vendors, utilities, regulators, owners' groups and other organisations such as WANO are considered The aerospace industry approach to Design Authority has been assessed to consider what lessons might be learned. (authors)

  1. Analyses in support of the Laboratory Microfusion Facility and ICF commercial reactor designs

    SciTech Connect (OSTI)

    Meier, W.R.; Monsler, M.J.

    1988-12-28

    Our work on this contract was divided into two major categories; two thirds of the total effort was in support of the Laboratory Microfusion Facility (LMF), and one third of the effort was in support of Inertial Confinement Fusion (ICF) commercial reactors. This final report includes copies of the formal reports, memoranda, and viewgraph presentations that were completed under this contract.

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    criteria for FBR/B&B safety systems/designs . . . . . . . . .Safety systems/designs violations of evaluation critera ARC-LL expansion liquid criteria . . . . . . . . . . . .criteria for systems and design-approaches to improve the inherent safety

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    Potential Safety Issues – Regulatory Design Criteria3-3 Regulatory design criteria for safety Table 3-4 Input3-4 Regulatory Design Criteria for safety The DRACS system

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

    E-Print Network [OSTI]

    Cisneros, Anselmo Tomas

    2013-01-01

    average   density   criteria   (design   sequence   1).  design  space,  design  criteria,  and  results  of  their  though  the  design   criteria-­?   and   optimization  

  5. Analytical Study on Thermal and Mechanical Design of Printed Circuit Heat Exchanger

    SciTech Connect (OSTI)

    Su-Jong Yoon; Piyush Sabharwall; Eung-Soo Kim

    2013-09-01

    The analytical methodologies for the thermal design, mechanical design and cost estimation of printed circuit heat exchanger are presented in this study. In this study, three flow arrangements of parallel flow, countercurrent flow and crossflow are taken into account. For each flow arrangement, the analytical solution of temperature profile of heat exchanger is introduced. The size and cost of printed circuit heat exchangers for advanced small modular reactors, which employ various coolants such as sodium, molten salts, helium, and water, are also presented.

  6. Fast Neutron Spectrum Potassium Worth for Space Power Reactor Design Validation

    SciTech Connect (OSTI)

    Bess, John D.; Marshall, Margaret A.; Briggs, J. Blair; Tsiboulia, Anatoli; Rozhikhin, Yevgeniy; Mihalczo, John T.

    2015-03-01

    A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, one was performed that consisted of uranium metal annuli surrounding a potassium-filled, stainless steel can. The outer diameter of the annuli was approximately 13 inches (33.02 cm) with an inner diameter of 7 inches (17.78 cm). The diameter of the stainless steel can was 7 inches (17.78 cm). The critical height of the configurations was approximately 5.6 inches (14.224 cm). The uranium annulus consisted of multiple stacked rings, each with radial thicknesses of 1 inch (2.54 cm) and varying heights. A companion measurement was performed using empty stainless steel cans; the primary purpose of these experiments was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs.The experimental measurements were performed on July 11, 1963, by J. T. Mihalczo and M. S. Wyatt (Ref. 1) with additional information in its corresponding logbook. Unreflected and unmoderated experiments with the same set of highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in the International Handbook for Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) with the identifier HEU MET FAST 051. Thin graphite reflected (2 inches or less) experiments also using the same set of highly enriched uranium metal parts are evaluated in HEU MET FAST 071. Polyethylene-reflected configurations are evaluated in HEU-MET-FAST-076. A stack of highly enriched metal discs with a thick beryllium top reflector is evaluated in HEU-MET-FAST-069, and two additional highly enriched uranium annuli with beryllium cores are evaluated in HEU-MET-FAST-059. Both detailed and simplified model specifications are provided in this evaluation. Both of these fast neutron spectra assemblies were determined to be acceptable benchmark experiments. The calculated eigenvalues for both the detailed and the simple benchmark models are within ~0.26 % of the benchmark values for Configuration 1 (calculations performed using MCNP6 with ENDF/B-VII.1 neutron cross section data), but under-calculate the benchmark values by ~7s because the uncertainty in the benchmark is very small: ~0.0004 (1s); for Configuration 2, the under-calculation is ~0.31 % and ~8s. Comparison of detailed and simple model calculations for the potassium worth measurement and potassium mass coefficient yield results approximately 70 – 80 % lower (~6s to 10s) than the benchmark values for the various nuclear data libraries utilized. Both the potassium worth and mass coefficient are also deemed to be acceptable benchmark experiment measurements.

  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)

    1999-01-01

    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. 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.

    1999-03-23

    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.

  9. Conversion of cellulosic wastes to liquid hydrocarbon fuels: Vol. 6, The modeling and design of a staged indirect liquefaction reactor: Final report

    SciTech Connect (OSTI)

    Kuester, J.L.

    1986-11-01

    A staged reactor was designed to convert biomass to useful fuels. The reactor consists of three stages. The first stage is a concentric combustor/pyrolyzer system where the biomass is gasified in a fluidized bed at high temperatures in the absence of oxygen. The second stage is a cyclonic scrubber where particulates and condensable materials are removed from the gas stream while the gas is cooled. In the final stage the gas undergoes a Fischer-Tropsch synthesis in a fluidized bed or slurry reactor. Mathematical models of the system were developed and used to create computer programs that would predict the behavior of the bed. The models were based on fundamental phenomena and were used to predict key dimensions of the staged reactor system. A transparent plastic, full-scale, cold flow reactor simulator was built using the models' predictions. The simulator was used to refine the models and determine the operating characteristics of the reactor. The design was determined to be workable and potentially useful. The reactor was, however, difficult to operate and would require extensive automated control systems.

  10. MODULAR MANIPULATOR FOR ROBOTICS APPLICATIONS

    SciTech Connect (OSTI)

    Joseph W. Geisinger, Ph.D.

    2001-07-31

    ARM Automation, Inc. is developing a framework of modular actuators that can address the DOE's wide range of robotics needs. The objective of this effort is to demonstrate the effectiveness of this technology by constructing a manipulator from these actuators within a glovebox for Automated Plutonium Processing (APP). At the end of the project, the system of actuators was used to construct several different manipulator configurations, which accommodate common glovebox tasks such as repackaging. The modular nature and quickconnects of this system simplify installation into ''hot'' boxes and any potential modifications or repair therein. This work focused on the development of self-contained robotic actuator modules including the embedded electronic controls for the purpose of building a manipulator system. Both of the actuators developed under this project contain the control electronics, sensors, motor, gear train, wiring, system communications and mechanical interfaces of a complete robotics servo device. Test actuators and accompanying DISC{trademark}s underwent validation testing at The University of Texas at Austin and ARM Automation, Inc. following final design and fabrication. The system also included custom links, an umbilical cord, an open architecture PC-based system controller, and operational software that permitted integration into a completely functional robotic manipulator system. The open architecture on which this system is based avoids proprietary interfaces and communication protocols which only serve to limit the capabilities and flexibility of automation equipment. The system was integrated and tested in the contractor's facility for intended performance and operations. The manipulator was tested using the full-scale equipment and process mock-ups. The project produced a practical and operational system including a quantitative evaluation of its performance and cost.

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

    E-Print Network [OSTI]

    Meagher, Paul Christopher

    1976-01-01

    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 ...

  12. High Efficiency Modular Chemical Processes (HEMCP)

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

    - ADVANCED MANUFACTURING OFFICE High Efficiency Modular Chemical Processes (HEMCP) Modular Process Intensification Framework for R&D Targets Advanced Manufacturing Office...

  13. 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

    2012-06-01

    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.

  14. Modular Quantum Cosmology

    E-Print Network [OSTI]

    Orfeu Bertolami; Ricardo Schiappa

    1999-04-30

    We study solutions of the Wheeler-DeWitt equation corresponding to an S-modular invariant N=1 supergravity model and a closed homogeneous and isotropic Friedmann-Robertson-Walker spacetime. The issues of inflation and of the cosmological constant problem are addressed with the help of the relevant wave functions. We find that topological type inflation is consistent from the quantum mechanical point of view and that a solution for the cosmological constant problem along the lines of the strong CP problem naturally arises.

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

    DOE Patents [OSTI]

    Wanlass, M.

    1985-02-19

    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.

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

    E-Print Network [OSTI]

    Merritt, Stanley Duane

    1991-01-01

    and organic oxygen. Qn a mass basis, lignites contain very little sulfur; but since the NSPS are based on a pounds of SQx per million Btu measurement, one finds that lignites are dirtier than the high medium volatile bituminous, mvb, coals of the Midwest... if it were not for a moisture content of approximately 30%. All coals were stored under nitrogen to prevent weathering and loss of moisture from the coal. Coal samples were weighed into glass vials for transport to the reactor. A blank was also prepared...

  17. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    Douglas M. Gerstner

    2009-05-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

  18. Integrating accident management issues in the design of future reactors EDF tentative approach

    SciTech Connect (OSTI)

    Berbey, P.; Vidard, M. [EDF-SEPTEN, Villeurbanne (France)

    1997-12-01

    In next generation plants, Severe Accidents will be explicitly considered at a very early stage of the design, and risk significant challenges will be addressed as appropriate. As design provisions could be considered to address some of these challenges, the need for Severe Accident Management (SAM) could be debated. For EDF, Accident Management (AM) in general, and SAM in particular, will remain a cornerstone of plant safety. Design provisions, if any, will have to be such as not to preclude any SAM measure with the potential for preventing accident sequences from progressing further. 3 refs., 2 figs.

  19. 11/04/02 G. W. Rubloff AVS 2002 MS MoA5 1 Spatially Programmable Reactor Design

    E-Print Network [OSTI]

    Rubloff, Gary W.

    technology nodes (roadmap) ­ Increasing demand for new, complex materials and processes · Manufacturing ­ Tuning for process integration (yield) ­ Technology shrink MANUFACTURING EFFICIENCY deposition rate Susceptor Wafer Reactor Heated Showerhead Heated Susceptor Wafer Gas Exhaust to Process Pumps Reactor A BCDE

  20. 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)

    2013-01-01

    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 ...

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

    E-Print Network [OSTI]

    Gezelius, Knut, 1978-

    2004-01-01

    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 ...

  2. 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; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik; Aziz, Ferhat; Sekimoto, Hiroshi

    2014-02-12

    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.

  3. Emergent spacetime from modular motives

    E-Print Network [OSTI]

    Rolf Schimmrigk

    2008-12-23

    The program of constructing spacetime geometry from string theoretic modular forms is extended to Calabi-Yau varieties of dimensions two, three, and four, as well as higher rank motives. Modular forms on the worldsheet can be constructed from the geometry of spacetime by computing the L-functions associated to omega motives of Calabi-Yau varieties, generated by their holomorphic $n-$forms via Galois representations. The modular forms that emerge from the omega motive and other motives of the intermediate cohomology are related to characters of the underlying rational conformal field theory. The converse problem of constructing space from string theory proceeds in the class of diagonal theories by determining the motives associated to modular forms in the category of motives with complex multiplication. The emerging picture indicates that the L-function can be interpreted as a map from the geometric category of motives to the category of conformal field theories on the worldsheet.

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    A. W. HUNE. “Accelerator Driven Systems for Transmutation:for use in accelerator-driven systems (ADS) as they can beAccelerator-driven B&Bs Terrapower LLC Commercial sodium-cooled SWR designs Figure 2.2: The research history of B&B systems

  5. Research and Development Assessments for Prometheus Heavy Ion and Laser Driven Inertial Fusion Energy Reactor Designs

    E-Print Network [OSTI]

    Tillack, Mark

    station electric power plants have been conceptually designed and analyzed in the Prometheus[1] study led by McDonnell Douglas Aerospace. These plants use inertial fusion energy (IFE) technologies by employing with a list of important R&D tasks that need to be conducted, and (3) identify areas of R&D that are common

  6. Safety approaches for high power modular laser operation

    SciTech Connect (OSTI)

    Handren, R.T.

    1993-03-01

    Approximately 20 years ago, a program was initiated at the Lawrence Livermore National Laboratory (LLNL) to study the feasibility of using lasers to separate isotopes of uranium and other materials. Of particular interest has been the development of a uranium enrichment method for the production of commercial nuclear power reactor fuel to replace current more expensive methods. The Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) Program has progressed to the point where a plant-scale facility to demonstrate commercial feasibility has been built and is being tested. The U-AVLIS Program uses copper vapor lasers which pump frequency selective dye lasers to photoionize uranium vapor produced by an electron beam. The selectively ionized isotopes are electrostatically collected. The copper lasers are arranged in oscillator/amplifier chains. The current configuration consists of 12 chains, each with a nominal output of 800 W for a system output in excess of 9 kW. The system requirements are for continuous operation (24 h a day, 7 days a week) and high availability. To meet these requirements, the lasers are designed in a modular form allowing for rapid change-out of the lasers requiring maintenance. Since beginning operation in early 1985, the copper lasers have accumulated over 2 million unit hours at a >90% availability. The dye laser system provides approximately 2.5 kW average power in the visible wavelength range. This large-scale laser system has many safety considerations, including high-power laser beams, high voltage, and large quantities ({approximately}3000 gal) of ethanol dye solutions. The Laboratory`s safety policy requires that safety controls be designed into any process, equipment, or apparatus in the form of engineering controls. Administrative controls further reduce the risk to an acceptable level. Selected examples of engineering and administrative controls currently being used in the U-AVLIS Program are described.

  7. Catalytic reactor

    DOE Patents [OSTI]

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

    2009-03-10

    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.

  8. An autonomous long-term fast reactor system and the principal design limitations of the concept 

    E-Print Network [OSTI]

    Tsvetkova, Galina Valeryevna

    2004-09-30

    VALERYEVNA TSVETKOVA Submitted to Texas A&M University in partial fulfillment of the requirements for the degree of DOCTOR OF PHILOSOPHY Approved as to style and content by: Kenneth L. Peddicord (Chair of Committee) Yassin A... System and the Principal Design Limitations of the Concept. (December 2003) Galina Valeryevna Tsvetkova, Dipl., Moscow State Engineering Physics Institute,Russia Chair of Advisory Committee: Dr. Kenneth L. Peddicord The objectives of this dissertation...

  9. Modular container assembled from fiber reinforced thermoplastic sandwich panels

    DOE Patents [OSTI]

    Donnelly, Mathew William (Edgewood, NM); Kasoff, William Andrew (Albuquerque, NM); Mcculloch, Patrick Carl (Irvine, CA); Williams, Frederick Truman (Albuquerque, NM)

    2007-12-25

    An improved, load bearing, modular design container structure assembled from thermoformed FRTP sandwich panels in which is utilized the unique core-skin edge configuration of the present invention in consideration of improved load bearing performance, improved useful load volume, reduced manufacturing costs, structural weight savings, impact and damage tolerance and repair and replace issues.

  10. A modular description for collimator EGS simulation tasks

    E-Print Network [OSTI]

    Lanconelli, Nico

    A modular description for collimator EGS simulation tasks geometry in Alessandro Bevilacqua, Dante, Alessandro Riccardi Abstract-EGS is a very common Monte Carlo code, used in the simulation of Nuclear configuration and camera design in Single Photon Emission studies. Using the EGS code, users must define

  11. Modular low-aspect-ratio high-beta torsatron

    DOE Patents [OSTI]

    Sheffield, G.V.

    1982-04-01

    A fusion-reactor device is described which the toroidal magnetic field and at least a portion of the poloidal magnetic field are provided by a single set of modular coils. The coils are arranged on the surface of a low-aspect-ratio toroid in planed having the cylindrical coordinate relationship phi = phi/sub i/ + kz, where k is a constant equal to each coil's pitch and phi/sub i/ is the toroidal angle at which the i'th coil intersects the z = o plane. The toroid defined by the modular coils preferably has a race track minor cross section. When vertical field coils and, preferably, a toroidal plasma current are provided for magnetic-field-surface closure within the toroid, a vacuum magnetic field of racetrack-shaped minor cross section with improved stability and beta valves is obtained.

  12. 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

    1994-08-01

    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.

  13. 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

    1994-08-01

    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.

  14. A Comparative Study of Modular Axial Flux Podded Generators for Marine Current Turbines

    E-Print Network [OSTI]

    Brest, Université de

    A Comparative Study of Modular Axial Flux Podded Generators for Marine Current Turbines Sofiane turbines (MCTs). Due to the submarine environment, maintenance operations are very hard, very costly current turbine, axial flux permanent magnet generator, design, optimization. Nomenclature MCT = Marine

  15. Modular Lorentz force actuators for efficient biomimetic propulsion of Autonomous Underwater Vehicles

    E-Print Network [OSTI]

    Church, Joseph Christopher

    2014-01-01

    In this thesis, we developed a highly scalable design for modular Lorentz force actuators for use in segmented flexible-hull undersea vehicles such as the RoboTuna being developed at Franklin W, Olin College of Engineering. ...

  16. Multidimensional bioseparation with modular microfluidics | SciTech...

    Office of Scientific and Technical Information (OSTI)

    Multidimensional bioseparation with modular microfluidics Citation Details In-Document Search Title: Multidimensional bioseparation with modular microfluidics You are accessing...

  17. MODULAR CONSTRUCTION SYSTEM EVALUATION

    SciTech Connect (OSTI)

    S. Gillespie

    2002-08-08

    The purpose of this study is to respond to U.S. Department of Energy (DOE) Technical Direction Letter (TDL) 02-003 (Waisley 2001), which directs Bechtel SAIC Company, LLC (BSC) to complete a design study to recommend repository design options to support receipt and/or emplacement of any or all of the following: commercial spent nuclear fuel (CSNF), high-level radioactive waste (HLW), DOE-managed spent nuclear fuel (DSNF) (including naval spent nuclear fuel [SNF]), and immobilized plutonium (if available), as soon as practicable, but no later than 2010. From the possible design options, a recommended approach will be determined for further evaluation to support the preliminary design of the repository. This study integrates the results of the repository Design Evolution Study (Rowe 2002) with supporting studies concerning national transportation options (BSC 2002b) and Nevada transportation options (Gehner 2002). The repository Design Evolution Study documents the processes used to reevaluate the design, construction, operation, and cost of the repository in response to TDL 02-003 (Waisley 2001), and to determine possible repository conceptual design options. The transportation studies evaluate the national and Nevada transportation options that support the repository conceptual design options. An evaluation methodology was established, based on Program-level requirements developed for the study in reference BSC 2001a, to allow the repository and system design options to be evaluated on a consistent basis. The transportation options and the design components were integrated into system design implementation options, which were evaluated using receipt and emplacement scenarios. The scenarios tested the ability of the design concept to adapt to changes in funding, waste receipt rate, and Nevada rail transportation availability. The results of the evaluation (in terms of system throughput, cost, and schedule) were then compared to the Program-level requirements, and recommendations for design alternatives, requirements changes, or further evaluation were developed.

  18. Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors

    E-Print Network [OSTI]

    Zweibaum, Nicolas

    2015-01-01

    Systems for Advanced Nuclear Reactors,” Ph.D. Dissertation,Handbook for Nuclear Reactor Safety,” Luxembourg: Commissiondevelopment of advanced nuclear reactor technology requires

  19. In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

    SciTech Connect (OSTI)

    Upadhyaya, Belle; Hines, J. Wesley; Damiano, Brian; Mehta, Chaitanya; Collins, Price; Lish, Matthew; Cady, Brian; Lollar, Victor; de Wet, Dane; Bayram, Duygu

    2015-12-15

    The research and development under this project was focused on the following three major objectives: Objective 1: Identification of critical in-vessel SMR components for remote monitoring and development of their low-order dynamic models, along with a simulation model of an integral pressurized water reactor (iPWR). Objective 2: Development of an experimental flow control loop with motor-driven valves and pumps, incorporating data acquisition and on-line monitoring interface. Objective 3: Development of stationary and transient signal processing methods for electrical signatures, machinery vibration, and for characterizing process variables for equipment monitoring. This objective includes the development of a data analysis toolbox. The following is a summary of the technical accomplishments under this project: - A detailed literature review of various SMR types and electrical signature analysis of motor-driven systems was completed. A bibliography of literature is provided at the end of this report. Assistance was provided by ORNL in identifying some key references. - A review of literature on pump-motor modeling and digital signal processing methods was performed. - An existing flow control loop was upgraded with new instrumentation, data acquisition hardware and software. The upgrading of the experimental loop included the installation of a new submersible pump driven by a three-phase induction motor. All the sensors were calibrated before full-scale experimental runs were performed. - MATLAB-Simulink model of a three-phase induction motor and pump system was completed. The model was used to simulate normal operation and fault conditions in the motor-pump system, and to identify changes in the electrical signatures. - A simulation model of an integral PWR (iPWR) was updated and the MATLAB-Simulink model was validated for known transients. The pump-motor model was interfaced with the iPWR model for testing the impact of primary flow perturbations (upsets) on plant parameters and the pump electrical signatures. Additionally, the reactor simulation is being used to generate normal operation data and data with instrumentation faults and process anomalies. A frequency controller was interfaced with the motor power supply in order to vary the electrical supply frequency. The experimental flow control loop was used to generate operational data under varying motor performance characteristics. Coolant leakage events were simulated by varying the bypass loop flow rate. The accuracy of motor power calculation was improved by incorporating the power factor, computed from motor current and voltage in each phase of the induction motor.- A variety of experimental runs were made for steady-state and transient pump operating conditions. Process, vibration, and electrical signatures were measured using a submersible pump with variable supply frequency. High correlation was seen between motor current and pump discharge pressure signal; similar high correlation was exhibited between pump motor power and flow rate. Wide-band analysis indicated high coherence (in the frequency domain) between motor current and vibration signals. - Wide-band operational data from a PWR were acquired from AMS Corporation and used to develop time-series models, and to estimate signal spectrum and sensor time constant. All the data were from different pressure transmitters in the system, including primary and secondary loops. These signals were pre-processed using the wavelet transform for filtering both low-frequency and high-frequency bands. This technique of signal pre-processing provides minimum distortion of the data, and results in a more optimal estimation of time constants of plant sensors using time-series modeling techniques.

  20. Modular multivariable control improves hydrocracking

    SciTech Connect (OSTI)

    Chia, T.L.; Lefkowitz, I. [ControlSoft, Inc., Cleveland, OH (United States); Tamas, P.D. [Marathon Oil Co., Robinson, IL (United States)

    1996-10-01

    Modular multivariable control (MMC), a system of interconnected, single process variable controllers, can be a user-friendly, reliable and cost-effective alternative to centralized, large-scale multivariable control packages. MMC properties and features derive directly from the properties of the coordinated controller which, in turn, is based on internal model control technology. MMC was applied to a hydrocracking unit involving two process variables and three controller outputs. The paper describes modular multivariable control, MMC properties, tuning considerations, application at the DCS level, constraints handling, and process application and results.

  1. Analysis of granular flow in a pebble-bed nuclear reactor Chris H. Rycroft,1

    E-Print Network [OSTI]

    Bazant, Martin Z.

    Analysis of granular flow in a pebble-bed nuclear reactor Chris H. Rycroft,1 Gary S. Grest,2 James February 2006; published 24 August 2006 Pebble-bed nuclear reactor technology, which is currently being States, the Modular Pebble Bed Reactor MPBR 4,8 is a candidate for the next generation nuclear plant

  2. 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

    1994-01-01

    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.

  3. Radiation Shielding Design and Orientation Considerations for a 1 kWe Heat Pipe Cooled Reactor Utilized to Bore Through the Ice Caps of Mars

    SciTech Connect (OSTI)

    Fensin, Michael L.; Elliott, John O.; Lipinski, Ronald J.; Poston, David I.

    2006-01-20

    The goal in designing any space power system is to develop a system able to meet the mission requirements for success while minimizing the overall costs. The mission requirements for the this study was to develop a reactor (with Stirling engine power conversion) and shielding configuration able to fit, along with all the other necessary science equipment, in a Cryobot 3 m high with {approx}0.5 m diameter hull, produce 1 kWe for 5yrs, and not adversely affect the mission science by keeping the total integrated dose to the science equipment below 150 krad. Since in most space power missions the overall system mass dictates the mission cost, the shielding designs in this study incorporated Martian water extracted at the startup site in order to minimize the tungsten and LiH mass loading at launch. Different reliability and mass minimization concerns led to three design configuration evolutions. With the help of implementing Martian water and configuring the reactor as far from the science equipment as possible, the needed tungsten and LiH shield mass was minimized. This study further characterizes the startup dose and the necessary mission requirements in order to ensure integrity of the surface equipment during reactor startup phase.

  4. Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors

    E-Print Network [OSTI]

    Zweibaum, Nicolas

    2015-01-01

    test SFR – Sodium-cooled fast reactor SWU – Separative workvalues than sodium-cooled fast reactors ( SFRs) and HTGRs.

  5. DESIGNING COMMUNICATION NETWORKS VIA HILBERT MODULAR FORMS

    E-Print Network [OSTI]

    Network Theory, the arithmetic examples are partic- ularly interesting: all Ramanujan graphs are super-expanders; but in addition the examples have many other useful properties, for example very good expansion constants

  6. DESIGNING COMMUNICATION NETWORKS VIA HILBERT MODULAR FORMS

    E-Print Network [OSTI]

    Theory, the arithmetic examples are partic- ularly interesting: all Ramanujan graphs are super-expanders; but in addition the examples have many other useful properties, for example very good expansion constants

  7. Modular & Scalable Molten Salt Plant Design

    Broader source: Energy.gov [DOE]

    This presentation was delivered at the SunShot Concentrating Solar Power (CSP) Program Review 2013, held April 23–25, 2013 near Phoenix, Arizona.

  8. A modular programming language for engineering design

    E-Print Network [OSTI]

    Coffee, Thomas Merritt

    2008-01-01

    We introduce a new universal model of computation called MDPL that generalizes other functional models like the lambda calculus and combinatory logic. This model leads naturally to a new type of programming language that ...

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

    E-Print Network [OSTI]

    Buongiorno, Jacopo, 1971-

    2001-01-01

    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 ...

  10. 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 ...

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

    E-Print Network [OSTI]

    Pope, Michael A. (Michael Alexander)

    2006-01-01

    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 ...

  12. 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

    2010-01-01

    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 ...

  13. Spectral Modular Exponentiation Gokay Saldamli

    E-Print Network [OSTI]

    California at Davis, University of

    and exponentiation based on a new reduction oper- ation are proposed (Section 2). These methods work com- pletely to meet the asymptotic crossover of Sch¨onhage-Strassen, assuming the reduction has a constant cost describe a new method to perform the modular expo- nentiation operation, i.e., the computation of c = me

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

    SciTech Connect (OSTI)

    Chang 'Apollo', Chen

    2006-07-01

    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)

  15. Chemical Reactor Analysis and Optimal Digestion

    E-Print Network [OSTI]

    Jumars, Pete

    derived from basic principles o f chemical reactor analysis and design Deborah L. Penry and Peter in terms of chemical reactor components and then use principles of reactor design to identify variablesJ 310 Chemical Reactor Analysis and Optimal Digestion An optimal digestion theory can be readily

  16. Report by the Joint-Core Team for the Establishment of Technology Bases Required for the Development of a Fusion DEMO Reactor

    E-Print Network [OSTI]

    DEMO reactor, reactor design, technology bases, technological issues, roadmap, timeline, critical path

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

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01

    protection. SDCs are safety design criteria. For FHRs, thereand on meeting the FHR safety design criteria (SDCs). When9 Proposed FHR Safety Design Criteria (

  18. 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

    2002-10-01

    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.

  19. 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.

    2001-10-01

    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.

  20. Modular HTGR Safety Basis and Approach

    SciTech Connect (OSTI)

    Thomas Hicks

    2011-08-01

    The Next Generation Nuclear Plant (NGNP) will be a licensed commercial high temperature gas-cooled reactor (HTGR) capable of producing electricity and/or high temperature process heat for industrial markets supporting a range of end-user applications. The NGNP Project has adopted the 10 CFR 52 Combined License (COL) process, as recommended in the NGNP Licensing Strategy - A Report to Congress, dated August 2008, as the foundation for the NGNP licensing strategy [DOE/NRC 2008]. Nuclear Regulatory Commission (NRC) licensing of the NGNP plant utilizing this process will demonstrate the efficacy for licensing future HTGRs for commercial industrial applications. This information paper is one in a series of submittals that address key generic issues of the priority licensing topics as part of the process for establishing HTGR regulatory requirements. This information paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach with the NRC staff and public stakeholders. The NGNP project does not expect to receive comments on this information paper because other white papers are addressing key generic issues of the priority licensing topics in greater detail.

  1. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009

    SciTech Connect (OSTI)

    Chandler, David; Freels, James D; Ilas, Germina; Miller, James Henry; Primm, Trent; Sease, John D; Guida, Tracey; Jolly, Brian C

    2010-02-01

    This report documents progress made during FY 2009 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 reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

  2. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

    2009-03-01

    This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly 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 reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

  3. Modular Aneutronic Fusion Engine

    SciTech Connect (OSTI)

    Gary Pajer, Yosef Razin, Michael Paluszek, A.H. Glasser and Samuel Cohen

    2012-05-11

    NASA's JUNO mission will arrive at Jupiter in July 2016, after nearly five years in space. Since operational costs tend to rise with mission time, minimizing such times becomes a top priority. We present the conceptual design for a 10MW aneutronic fusion engine with high exhaust velocities that would reduce transit time for a Jupiter mission to eighteen months and enable more challenging exploration missions in the solar system and beyond. __________________________________________________

  4. Modular, multi-level groundwater sampler

    DOE Patents [OSTI]

    Nichols, Ralph L. (812 Plantation Point Dr., N. Augusta, SC 29841); Widdowson, Mark A. (4204 Havana Ct., Columbia, SC 29206); Mullinex, Harry (10 Cardross La., Columbia, SC 29209); Orne, William H. (12 Martha Ct., Sumter, SC 29150); Looney, Brian B. (1135 Ridgemont Dr., Aiken, SC 29803)

    1994-01-01

    Apparatus for taking a multiple of samples of groundwater or pressure measurements from a well simultaneously. The apparatus comprises a series of chambers arranged in an axial array, each of which is dimensioned to fit into a perforated well casing and leave a small gap between the well casing and the exterior of the chamber. Seals at each end of the container define the limits to the axial portion of the well to be sampled. A submersible pump in each chamber pumps the groundwater that passes through the well casing perforations into the gap from the gap to the surface for analysis. The power lines and hoses for the chambers farther down the array pass through each chamber above them in the array. The seals are solid, water-proof, non-reactive, resilient disks supported to engage the inside surface of the well casing. Because of the modular design, the apparatus provides flexibility for use in a variety of well configurations.

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

    SciTech Connect (OSTI)

    J. Wesley Hines; Belle R. Upadhyaya; J. Michael Doster; Robert M. Edwards; Kenneth D. Lewis; Paul Turinsky; Jamie Coble

    2011-05-31

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

  6. Mock Modular Mathieu Moonshine Modules

    E-Print Network [OSTI]

    Miranda C. N. Cheng; Xi Dong; John F. R. Duncan; Sarah Harrison; Shamit Kachru; Timm Wrase

    2015-08-10

    We construct super vertex operator algebras which lead to modules for moonshine relations connecting the four smaller sporadic simple Mathieu groups with distinguished mock modular forms. Starting with an orbifold of a free fermion theory, any subgroup of Co_0 that fixes a 3-dimensional subspace of its unique non-trivial 24-dimensional representation commutes with a certain N=4 superconformal algebra. Similarly, any subgroup of Co_0 that fixes a 2-dimensional subspace of the 24-dimensional representation commutes with a certain N=2 superconformal algebra. Through the decomposition of the corresponding twined partition functions into characters of the N=4 (resp. N=2) superconformal algebra, we arrive at mock modular forms which coincide with the graded characters of an infinite-dimensional Z-graded module for the corresponding group. The Mathieu groups are singled out amongst various other possibilities by the moonshine property: requiring the corresponding weak Jacobi forms to have certain asymptotic behaviour near cusps. Our constructions constitute the first examples of explicitly realized modules underlying moonshine phenomena relating mock modular forms to sporadic simple groups. Modules for other groups, including the sporadic groups of McLaughlin and Higman--Sims, are also discussed.

  7. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    SciTech Connect (OSTI)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  8. 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)

    2014-10-06

    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.

  9. 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

    2013-11-29

    This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and 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.

  10. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    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)

  11. Multidimensional bioseparation with modular microfluidics Chirica...

    Office of Scientific and Technical Information (OSTI)

    Multidimensional bioseparation with modular microfluidics Chirica, Gabriela S.; Renzi, Ronald F. A multidimensional chemical separation and analysis system is described including a...

  12. Rank-finiteness for modular categories

    E-Print Network [OSTI]

    Paul Bruillard; Siu-Hung Ng; Eric C. Rowell; Zhenghan Wang

    2015-05-27

    We prove a rank-finiteness conjecture for modular categories: up to equivalence, there are only finitely many modular categories of any fixed rank. Our technical advance is a generalization of the Cauchy theorem in group theory to the context of spherical fusion categories. For a modular category $\\mathcal{C}$ with $N=ord(T)$, the order of the modular $T$-matrix, the Cauchy theorem says that the set of primes dividing the global quantum dimension $D^2$ in the Dedekind domain $\\mathbb{Z}[e^{\\frac{2\\pi i}{N}}]$ is identical to that of $N$.

  13. Evaluation of Potential Locations for Siting Small Modular Reactors near

    Energy Savers [EERE]

    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 on DeliciousMathematicsEnergy HeadquartersFuelB IMSof 2005 atHouse Environmental1-01 EvaluationOE-2: 2015-1

  14. Electricity Generating Portfolios with Small Modular Reactors | Department

    Energy Savers [EERE]

    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 on DeliciousMathematicsEnergy HeadquartersFuelB IMS Monthly Review EVMSand Transcript |Federal

  15. Small Modular Reactors, National Security and Clean Energy: A...

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

    and proliferation." http:mediacentral.princeton.eduid07bhf9ilf Contact Information Coordinator(s): Carol Ann Austin caustin@pppl.gov Host(s): Dr. Adam B. Cohen acohen@pppl.gov...

  16. Performance and Safety Analysis of a Generic Small Modular Reactor 

    E-Print Network [OSTI]

    Kitcher, Evans Damenortey, 1987-

    2012-11-07

    a fuel assembly position, each white square 150 200 250 300 350 400 35 40 45 50 55 60 65 70 75 80 Act iv e F uel Leng th Number of Assemblies Number of Assemblies required for the core 17 representing water filled positions... performance criteria of the mPower SMR from B&W. The Monte Carlo codes MCNP5/MCNPX are used to model the core. Fuel enrichment, core inventory, core size are all variables optimized to meet the set goals of core lifetime and fuel utilization (burnup...

  17. Economic Analysis of the Modular Pebble Bed Reactor

    E-Print Network [OSTI]

    of scale vs. productivity #12;MPBR Cost Estimate · Capital cost · O&M cost · Fuel cost · Decommissioning cost #12;Capital Cost · Cost savings come from: ­ more factory fabrication, less site work ­ learning effect from 1st to 10th unit ­ natural safety features ­ shorter construction time · Total capital cost

  18. Benefits of Small Modular Reactors (SMRs) | Department of Energy

    Energy Savers [EERE]

    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 on DeliciousMathematics And StatisticsProgram ManagerCorridor6 Audit Report: WR-B-00-06B l u e RPersistent,Benefits of

  19. Partnerships Help Advance Small Modular Reactor Technology | Department of

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

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergy AEnergyPresidential PermitDAYS - WE NEED A CHANGEof The Department of

  20. Development and Optimization of Modular Hybrid Plasma Reactor (Technical

    Office of Scientific and Technical Information (OSTI)

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield MunicipalTechnical Report: Achievements ofCOMPOSITION OF VAPORSSeries)SupportedDavidA - LReport) |

  1. Development and Optimization of Modular Hybrid Plasma Reactor (Technical

    Office of Scientific and Technical Information (OSTI)

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfate Reducing BacteriaConnectlaser-solidSwitchgrass and Miscanthus x GiganteusReport) | SciTech

  2. Economic Aspects of Small Modular Reactors | Department of Energy

    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 on Delicious Rank EERE:FinancingPetroleum Based|DepartmentStatementofApril 25,EV Everywhere|Muscle Car |EcoCAR: The

  3. Electricity Generating Portfolios with Small Modular Reactors | 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 on Delicious Rank EERE:FinancingPetroleum Based|DepartmentStatementofAprilof Energy 2ofU.S. Department ofEnergyof

  4. Small Modular Reactors Presentation to Secretary of Energy Advisory Board -

    Energy Savers [EERE]

    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 on QA:QA J-E-1 SECTION J APPENDIX E LIST OF APPLICABLE DIRECTIVES PursuantEnergy Small Column Ion| Department ofDeputy

  5. Energy Department Announces Small Modular Reactor Technology Partnerships

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

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirley Ann JacksonDepartment ofOffice ofofWind Projects | DepartmentRoofs, and Heating and Coolingat

  6. 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)

    2009-11-17

    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.

  7. Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

    SciTech Connect (OSTI)

    Hursin, M.; Perret, G. [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland)

    2012-07-01

    The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

  8. Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts

    SciTech Connect (OSTI)

    Ronald Farris; David Gertman; Jacques Hugo

    2014-03-01

    This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was to develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that would provide sophisticated operational information visualization, coupled with adaptive automation schemes and operator support systems to reduce complexity. These all have to be mapped at some point to human performance requirements. The EBR-II results will be used as a baseline that will be extrapolated in the extended Cognitive Work Analysis phase to the analysis of a selected advanced sodium-cooled SMR design as a way to establish non-conventional operational concepts. The Work Domain Analysis results achieved during this phase have not only established an organizing and analytical framework for describing existing sociotechnical systems, but have also indicated that the method is particularly suited to the analysis of prospective and immature designs. The results of the EBR-II Work Domain Analysis have indicated that the methodology is scientifically sound and generalizable to any operating environment.

  9. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    SciTech Connect (OSTI)

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  10. A hybrid network architecture for modular data centers

    E-Print Network [OSTI]

    Radhakrishnan, Sivasankar

    2010-01-01

    Part-time Optics in Data Centers. In ACM SIGCOMM ’10. [WLLStructure for Modular Data Center Interconnection. In ACMArchitecture for Modular Data Centers A thesis submitted in

  11. Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDU{sup R} and ACR{sup TM} reactors

    SciTech Connect (OSTI)

    Aydogdu, K.; Boss, C. R. [Atomic Energy of Canada Limited, Sheridan Science and Technology Park, Mississauga, Ont. L5K 1B2 (Canada)

    2006-07-01

    This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies heavily on experience and engineering judgement, consistent with the ALARA philosophy. Special care is taken to ensure that the best estimate dose rates are used to the extent possible when applying ALARA. Provisions for safeguards equipment are made throughout the fuel-handling route in CANDU and ACR reactors. For example, the fuel bundle counters rely on the decay gammas from the fission products in spent-fuel bundles to record the number of fuel movements. The International Atomic Energy Agency (IAEA) Safeguards system for CANDU and ACR reactors is based on item (fuel bundle) accounting. It involves a combination of IAEA inspection with containment and surveillance, and continuous unattended monitoring. The spent fuel bundle counter monitors spent fuel bundles as they are transferred from the fuelling machine to the spent fuel bay. The shielding and dose-rate analysis need to be carried out so that the bundle counter functions properly. This paper includes two codes used in criticality safety analyses. Criticality safety is a unique phenomenon and codes that address criticality issues will demand specific validations. However, it is recognised that some of the codes used in radiation physics will also be used in criticality safety assessments. (authors)

  12. An inverse of the modular invariant

    E-Print Network [OSTI]

    Semjon Adlaj

    2011-10-14

    During the last few years of his life, Ramanujan had adamantly tried to invert the modular invariant. Subsequent efforts failed until May 30, 2011 when an explicit closed formula for an inverse was presented at the CCRAS (Moscow, Russia). This very formula, along with some special values of the modular invariant, is given in this paper.

  13. Modularity of Termination Using Dependency Pairs ?

    E-Print Network [OSTI]

    Ábrahám, Erika

    Modularity of Termination Using Dependency Pairs ? Thomas Arts 1 and J¨urgen Giesl 2 1 Computer@informatik.tu­darmstadt.de Abstract. The framework of dependency pairs allows automated ter­ mination and innermost termination proofs of this framework in order to prove termination in a modular way. Our mod­ ularity results significantly increase

  14. On the Generator of Massive Modular Groups

    E-Print Network [OSTI]

    Timor Saffary

    2006-01-19

    The purpose of this paper is to shed more light on the transition from the known massless modular action to the wanted massive one in the case of forward light cones and double cones. The infinitesimal generator of the massive modular automorphism group is investigated, in particular, some assumptions on its structure are verified explicitly for two concrete examples.

  15. Efficient Software Implementations of Modular Exponentiation

    E-Print Network [OSTI]

    International Association for Cryptologic Research (IACR)

    1 Efficient Software Implementations of Modular Exponentiation Shay Gueron 1, 2 1 Department on the workloads of SSL/TLS servers, and therefore their software implementations on general purpose processors attacks. Together, these lead to an efficient software implementation of 512-bit modular exponentiation

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

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01

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

  17. NUHOMS modular spent-fuel storage system: Performance testing

    SciTech Connect (OSTI)

    Strope, L.A.; McKinnon, M.A. ); Dyksterhouse, D.J.; McLean, J.C. )

    1990-09-01

    This report documents the results of a heat transfer and shielding performance evaluation of the NUTECH HOrizontal MOdular Storage (NUHOMS{reg sign}) System utilized by the Carolina Power and Light Co. (CP L) in an Independent Spent Fuel Storage Installation (ISFSI) licensed by the US Nuclear Regulatory Commission (NRC). The ISFSI is located at CP L's H. B. Robinson Nuclear Plant (HBR) near Hartsville, South Carolina. The demonstration included testing of three modules, first with electric heaters and then with spent fuel. The results indicated that the system was conservatively designed, with all heat transfer and shielding design criteria easily met. 5 refs., 45 figs., 9 tabs.

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

    SciTech Connect (OSTI)

    Thomas E. Hicks

    2011-07-01

    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.

  19. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03

    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.

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

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01

    Design. (U.S. Nuclear Regulatory Commission, NUREG-1793,analysis. (U.S. Nuclear Regulatory Commission, 1991, NUREG/Main Report. (Nuclear Regulatory Commission, 2007, NUREG/CR-

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

    SciTech Connect (OSTI)

    Adeniyi Lawal

    2008-12-09

    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.

  2. Microfluidic reactors for the synthesis of nanocrystals

    E-Print Network [OSTI]

    Yen, Brian K. H

    2007-01-01

    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 ...

  3. 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 ...

  4. 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)

    2005-01-01

    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 ...

  5. 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 ...

  6. 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.)

    1982-02-09

    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.

  7. Modular, multi-level groundwater sampler

    DOE Patents [OSTI]

    Nichols, R.L.; Widdowson, M.A.; Mullinex, H.; Orne, W.H.; Looney, B.B.

    1994-03-15

    An apparatus is described for taking a multiple of samples of groundwater or pressure measurements from a well simultaneously. The apparatus comprises a series of chambers arranged in an axial array, each of which is dimensioned to fit into a perforated well casing and leave a small gap between the well casing and the exterior of the chamber. Seals at each end of the container define the limits to the axial portion of the well to be sampled. A submersible pump in each chamber pumps the groundwater that passes through the well casing perforations into the gap from the gap to the surface for analysis. The power lines and hoses for the chambers farther down the array pass through each chamber above them in the array. The seals are solid, water-proof, non-reactive, resilient disks supported to engage the inside surface of the well casing. Because of the modular design, the apparatus provides flexibility for use in a variety of well configurations. 3 figures.

  8. Modular Electric Vehicle Program (MEVP). Final technical report

    SciTech Connect (OSTI)

    1994-03-01

    The Modular Electric Vehicle Program (MEVP) was an EV propulsion system development program in which the technical effort was contracted by DOE to Ford Motor Company. The General Electric Company was a major subcontractor to Ford for the development of the electric subsystem. Sundstrand Power Systems was also a subcontractor to Ford, providing a modified gas turbine engine APU for emissions and performance testing as well as a preliminary design and producibility study for a Gas Turbine-APU for potential use in hybrid/electric vehicles. The four-year research and development effort was cost-shared between Ford, General Electric, Sundstrand Power Systems and DOE. The contract was awarded in response to Ford`s unsolicited proposal. The program objective was to bring electric vehicle propulsion system technology closer to commercialization by developing subsystem components which can be produced from a common design and accommodate a wide range of vehicles; i.e., modularize the components. This concept would enable industry to introduce electric vehicles into the marketplace sooner than would be accomplished via traditional designs in that the economies of mass production could be realized across a spectrum of product offerings. This would eliminate the need to dedicate the design and capital investment to a limited volume product offering which would increase consumer cost and/or lengthen the time required to realize a return on the investment.

  9. 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

    2014-03-01

    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.

  10. Modular Power Converters for PV Applications

    SciTech Connect (OSTI)

    Ozpineci, Burak; Tolbert, Leon M

    2012-05-01

    This report describes technical opportunities to serve as parts of a technological roadmap for Shoals Technologies Group in power electronics for PV applications. There are many different power converter circuits that can be used for solar inverter applications. The present applications do not take advantage of the potential for using common modules. We envision that the development of a power electronics module could enable higher reliability by being durable and flexible. Modules would have fault current limiting features and detection circuits such that they can limit the current through the module from external faults and can identify and isolate internal faults such that the remaining modules can continue to operate with only minimal disturbance to the utility or customer. Development of a reliable, efficient, low-cost, power electronics module will be a key enabling technology for harnessing more power from solar panels and enable plug and play operation. Power electronics for computer power supplies, communication equipment, and transportation have all targeted reliability and modularity as key requirements and have begun concerted efforts to replace monolithic components with collections of common smart modules. This is happening on several levels including (1) device level with intelligent control, (2) functional module level, and (3) system module. This same effort is needed in power electronics for solar applications. Development of modular units will result in standard power electronic converters that will have a lower installed and operating cost for the overall system. These units will lead to increased adaptability and flexibility of solar inverters. Incorporating autonomous fault current limiting and reconfiguration capabilities into the modules and having redundant modules will lead to a durable converter that can withstand the rigors of solar power generation for more than 30 years. Our vision for the technology roadmap is that there is no need for detailed design of new power converters for each new application or installation. One set of modules and controllers can be pre-developed and the only design question would be how many modules need to be in series or parallel for the specific power requirement. Then, a designer can put the modules together and add the intelligent reconfigurable controller. The controller determines how many modules are connected, but it might also ask for user input for the specific application during setup. The modules include protection against faults and can reset it, if necessary. In case of a power device failure, the controller reconfigures itself to continue limited operation until repair which might be as simple as taking the faulty module out and inserting a new module. The result is cost savings in design, maintenance, repair, and a grid that is more reliable and available. This concept would be a perfect fit for the recently announced funding opportunity announcement (DE-FOA-0000653) on Plug and Play Photovoltaics.

  11. Dismantling Structures and Equipment of the MR Reactor and its Loop Facilities at the National Research Center 'Kurchatov Institute' - 12051

    SciTech Connect (OSTI)

    Volkov, V.G.; Danilovich, A.S.; Zverkov, Yu. A.; Ivanov, O.P.; Kolyadin, V.I.; Lemus, A.V.; Muzrukova, V.D.; Pavlenko, V.I.; Semenov, S.G.; Fadin, S.Yu.; Shisha, A.D.; Chesnokov, A.V.

    2012-07-01

    In 2008 a design of decommissioning of research reactors MR and RFT has been developed in the National research Center 'Kurchatov institute'. The design has been approved by Russian State Authority in July 2009 year and has received the positive conclusion of ecological expertise. In 2009-2010 a preparation for decommissioning of reactors MR and RFT was spent. Within the frames of a preparation a characterization, sorting and removal of radioactive objects, including the irradiated fuel, from reactor storage facilities and pool have been executed. During carrying out of a preparation on removal of radioactive objects from reactor sluice pool water treating has been spent. For these purposes modular installation for clearing and processing of a liquid radioactive waste 'Aqua - Express' was used. As a result of works it was possible to lower volume activity of water on three orders in magnitude that has allowed improving essentially of radiating conditions in a reactor hall. Auxiliary systems of ventilation, energy and heat supplies, monitoring systems of radiating conditions of premises of the reactor and its loop-back installations are reconstructed. In 2011 the license for a decommissioning of the specified reactors has been received and there are begun dismantling works. Within the frames of works under the design the armature and pipelines are dismantled in a under floor space of a reactor hall where a moving and taking away pipelines of loop facilities and the first contour of the MR reactor were replaced. A dismantle of the main equipment of loop facility with the gas coolant has been spent. Technologies which were used on dismantle of the radioactive contaminated equipment are presented, the basic works on reconstruction of systems of maintenance of on the decommissioning works are described, the sequence of works on the decommissioning of reactors MR and RFT is shown. Dismantling works were carried out with application of means of a dust suppression that, in aggregate with standard means at such works of individual protection of the personnel and devices of radiating control, has allowed to lower risk of action of radiation on the personnel, the population and environment at the expense of reduction of volume activity of radioactive aerosols in air. (authors)

  12. 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

    2014-07-10

    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.

  13. Modular bioreactor for the remediation of liquid streams and methods for using the same

    DOE Patents [OSTI]

    Noah, K.S.; Sayer, R.L.; Thompson, D.N.

    1998-06-30

    The present invention is directed to a bioreactor system for the remediation of contaminated liquid streams. The bioreactor system is composed of at least one and often a series of sub-units referred to as bioreactor modules. The modular nature of the system allows bioreactor systems be subdivided into smaller units and transported to waste sites where they are combined to form bioreactor systems of any size. The bioreactor modules further comprises reactor fill materials in the bioreactor module that remove the contaminants from the contaminated stream. To ensure that the stream thoroughly contacts the reactor fill materials, each bioreactor module comprises means for directing the flow of the stream in a vertical direction and means for directing the flow of the stream in a horizontal direction. In a preferred embodiment, the reactor fill comprises a sulfate reducing bacteria which is particularly useful for precipitating metals from acid mine streams. 6 figs.

  14. Modular software architecture for flexible reservation mechanisms on heterogeneous Michal Sojka,a

    E-Print Network [OSTI]

    Lipari, Giuseppe

    counting, monitoring of parking lots, etc.). In industrial con- trol, image recognition applicationsModular software architecture for flexible reservation mechanisms on heterogeneous resources Michal. However, it is not easy to design basic infrastructure services that allow for an easy access

  15. Modular Full-System Verification of Hardware Muralidaran Vijayaraghavan and Joonwon Choi

    E-Print Network [OSTI]

    Modular Full-System Verification of Hardware Muralidaran Vijayaraghavan and Joonwon Choi 1 of full-system verification for hardware designs. What does a full-system verification for a hardware of how to give a hardware specification and its implementation, and what the respective semantics are

  16. 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.

    2007-12-15

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

  17. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, Roger B. (Penn Twp., PA); Fero, Arnold H. (New Kensington, PA); Sejvar, James (Murrysville, PA)

    1997-01-01

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  18. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

  19. Ssessment methodology for proliferation resistant fast breeder reactor

    E-Print Network [OSTI]

    Singh, Mohit, S.M. Massachusetts Institute of Technology

    2014-01-01

    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 ...

  20. 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.

    2003-12-01

    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

  1. Interfacial effects in fast reactors

    E-Print Network [OSTI]

    Saidi, Mohammad Said

    1979-01-01

    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 ...

  2. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    SciTech Connect (OSTI)

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions between the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very effective in removing decay heat. By removing the limit on the decay heat removal capability due to the limited available surface area as in a RVACS, the reactor power and power density can be significantly increased, without losing the passive heat removal feature. This paper will introduce the concept of using DRACS to enhance VHTR passive safety and economics. Three design options will be discussed, depending on the cooling pipe locations. Analysis results from a lumped volume based model and CFD simulations will be presented.

  3. Summary of the Preliminary Optical ICHMI Design Study: A Preliminary Engineering Design Study for a Standpipe Viewport

    SciTech Connect (OSTI)

    Anheier, Norman C.; Qiao, Hong; Berglin, Eric J.; Hatchell, Brian K.

    2013-12-26

    This summary report examines an in-vessel optical access concept intended to support standoff optical instrumentation, control and human-machine interface (ICHMI) systems for future advanced small modular reactor (AdvSMR) applications. Optical-based measurement and sensing systems for AdvSMR applications have several key benefits over traditional instrumentation and control systems used to monitor reactor process parameters, such as temperature, flow rate, pressure, and coolant chemistry (Anheier et al. 2013). Direct and continuous visualization of the in-vessel components can be maintained using external cameras. Many optical sensing techniques can be performed remotely using open optical beam path configurations. Not only are in-vessel cables eliminated by these configurations, but also sensitive optical monitoring components (e.g., electronics, lasers, detectors, and cameras) can be placed outside the reactor vessel in the instrument vault, containment building, or other locations where temperatures and radiation levels are much lower. However, the extreme AdvSMR environment present challenges for optical access designs and optical materials. Optical access is not provided in any commercial nuclear power plant or featured in any reactor design, although successful implementation of optical access has been demonstrated in test reactors (Arkani and Gharib 2009). This report outlines the key engineering considerations for an AdvSMR optical access concept. Strict American Society of Mechanical Engineers (ASME) construction codes must be followed for any U.S. nuclear facility component (ASME 2013); however, the scope of this study is to evaluate the preliminary engineering issues for this concept, rather than developing a nuclear-qualified design. In addition, this study does not consider accident design requirements. In-vessel optical access using a standpipe viewport concept serves as a test case to explore the engineering challenges and performance requirements for sodium fast reactor (SFR) and high-temperature gas reactor (HTGR) AdvSMR applications. The expected environmental conditions for deployment are reviewed for both AdvSMR designs. Optical and mechanical materials that maximize component lifetime are evaluated for the standpipe viewport design under these conditions. Optical components and opto-mechanical designs that provide robust optical-to-metal seals and stress-free optical component mounting are identified, and then key performance specifications are developed for a sapphire optical viewport concept. Design strategies are examined that protect the internal optical surfaces from liquid-coolant condensation and impurity deposits. Finally, a conceptual standpipe viewport design that is suggestive of how this concept could be assembled using standard nuclear-qualified pipe components, is presented.

  4. Fuel Development For Gas-Cooled Fast Reactors

    SciTech Connect (OSTI)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  5. MODULAR MAGIC SUDOKU JOHN LORCH AND ELLEN WELD

    E-Print Network [OSTI]

    Lorch, John D.

    MODULAR MAGIC SUDOKU JOHN LORCH AND ELLEN WELD Abstract. A modular magic sudoku solution. 05B15. 1 #12;2 JOHN LORCH AND ELLEN WELD sudoku solution in (1) and the solution x2 given in Section

  6. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    SciTech Connect (OSTI)

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  7. 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

    2012-07-16

    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.

  8. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    SciTech Connect (OSTI)

    Corradin, Michael; Anderson, M.; Muci, M.; Hassan, Yassin; Dominguez, A.; Tokuhiro, Akira; Hamman, K.

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  9. Integrated design : a generative multi-performative design approach

    E-Print Network [OSTI]

    Fasoulaki, Eleftheria

    2008-01-01

    There are building systems, called "modularized", in which the component systems (for structure, lighting, etc) can be analyzed and synthesized independently since their performance and design do not interact or affect one ...

  10. Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis

    SciTech Connect (OSTI)

    Wilson, Paul; Evans, Thomas; Tautges, Tim

    2012-12-24

    This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well-suited to coupling with the unstructured meshes that are used in other physics simulations.

  11. 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)

    1994-03-01

    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.

  12. Compact Reactor

    SciTech Connect (OSTI)

    Williams, Pharis E. [Williams Research, P.O. Box 554, Los Alamos, NM87544 (United States)

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  13. Aalborg Universitet Modular Power Architectures for Microgrid Clusters

    E-Print Network [OSTI]

    Vasquez, Juan Carlos

    Aalborg Universitet Modular Power Architectures for Microgrid Clusters Lin, Hengwei; Liu, Chengxi., Vasquez, J. C., & Dragicevic, T. (2014). Modular Power Architectures for Microgrid Clusters from vbn.aau.dk on: juli 04, 2015 #12;Modular Power Architectures for Microgrid Clusters ­ Invited

  14. A Modular Control System for Remote Subsea Eric Stephen Smith

    E-Print Network [OSTI]

    Wood, Stephen L.

    A Modular Control System for Remote Subsea Equipment by Eric Stephen Smith Bachelor of Science the undersigned committee hereby approve the attached thesis A Modular Control System for Remote Subsea Equipment and Environmental Systems #12;iv Abstract Title: A Modular Control System for Remote Subsea Equipment Author: Eric

  15. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul (Pittsburgh, PA)

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  16. 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; neutron diffusion and moderation; reactor equations; Fermi Age theory; multigroup and multiregional students should be able to: · Compare and contrast numerous nuclear reactor designs · Calculate the effects

  17. Semi-finished modular cells

    E-Print Network [OSTI]

    Bachelder, Laura Govoni, 1971-

    2002-01-01

    This thesis subject is a pre-fabricated element (cell): a system that employs natural, light, and economic materials to produce a near-finished portion of a building. The intent is to introduce sustainable design into ...

  18. Results from the DOE Advanced Gas Reactor Fuel Development and Qualification Program

    SciTech Connect (OSTI)

    David Petti

    2014-06-01

    Modular HTGR designs were developed to provide natural safety, which prevents core damage under all design basis accidents and presently envisioned severe accidents. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude and allows potential elimination of the need for evacuation and sheltering beyond a small exclusion area. This level, however, is predicated on exceptionally high fuel fabrication quality and performance under normal operation and accident conditions. Germany produced and demonstrated high quality fuel for their pebble bed HTGRs in the 1980s, but no U.S. manufactured fuel had exhibited equivalent performance prior to the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The design goal of the modular HTGRs is to allow elimination of an exclusion zone and an emergency planning zone outside the plant boundary fence, typically interpreted as being about 400 meters from the reactor. To achieve this, the reactor design concepts require a level of fuel integrity that is better than that claimed for all prior US manufactured TRISO fuel, by a few orders of magnitude. The improved performance level is about a factor of three better than qualified for German TRISO fuel in the 1980’s. At the start of the AGR program, without a reactor design concept selected, the AGR fuel program selected to qualify fuel to an operating envelope that would bound both pebble bed and prismatic options. This resulted in needing a fuel form that could survive at peak fuel temperatures of 1250°C on a time-averaged basis and high burnups in the range of 150 to 200 GWd/MTHM (metric tons of heavy metal) or 16.4 to 21.8% fissions per initial metal atom (FIMA). Although Germany has demonstrated excellent performance of TRISO-coated UO2 particle fuel up to about 10% FIMA and 1150°C, UO2 fuel is known to have limitations because of CO formation and kernel migration at the high burnups, power densities, temperatures, and temperature gradients that may be encountered in the prismatic modular HTGRs. With uranium oxycarbide (UCO) fuel, the kernel composition is engineered to prevent CO formation and kernel migration, which are key threats to fuel integrity at higher burnups, temperatures, and temperature gradients. Furthermore, the recent poor fuel performance of UO2 TRISO fuel pebbles measured in Chinese irradiation testing in Russia and in German pebbles irradiated at 1250°C, and historic data on poorer fuel performance in safety testing of German pebbles that experienced burnups in excess of 10% FIMA [1] have each raised concern about the use of UO2 TRISO above 10% FIMA and 1150°C and the degree of margin available in the fuel system. This continues to be an active area of study internationally.

  19. 1 INTRODUCTION The modular finitedifference groundwater flow

    E-Print Network [OSTI]

    Russell, Thomas F.

    1 INTRODUCTION The modular finite­difference ground­water flow model (MODFLOW) developed by the U­dimensional ground­water systems (McDonald & Harbaugh, 1988, Harbaugh & McDonald, 1996). MOC3D is a solute is optimal for advection­ dominated systems, which are typical of many field problems involving ground­water

  20. Improved Modular Termination Proofs Using Dependency Pairs

    E-Print Network [OSTI]

    Middeldorp, Aart

    Improved Modular Termination Proofs Using Dependency Pairs Ren´e Thiemann, J¨urgen Giesl, Peter) termination proofs of term rewrite systems (TRSs). For any TRS, it generates inequality constraints that have to be satisfied by well-founded orders. However, proving innermost termination is considerably easier than

  1. Improved Modular Termination Proofs Using Dependency Pairs

    E-Print Network [OSTI]

    Ábrahám, Erika

    Improved Modular Termination Proofs Using Dependency Pairs Renâ??e Thiemann, JË?urgen Giesl, Peter) termination proofs of term rewrite systems (TRSs). For any TRS, it generates inequality constraints that have to be satisfied by well­founded orders. However, proving innermost termination is considerably easier than

  2. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    SciTech Connect (OSTI)

    David E. Shropshire

    2004-04-01

    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.

  3. Final Technical Report for the Period September 2002 through September 2005; H2-MHR Pre-Conceptual Design Report: SI-Based Plant; H2-MHR Pre-Conceptual Design Report: HTE-Based Plant

    SciTech Connect (OSTI)

    M. Richards; A. Shenoy; L. Brown; R. Buckingham; E. Harvego; K. Peddicord; M. Reza; J. Coupey

    2006-04-19

    For electricity and hydrogen production, an advanced reactor technology receiving considerable international interest is a modular, passively-safe version of the high-temperature, gas-cooled reactor, known in the U.S. as the Modular Helium Reactor (MHR), which operates at a power level of 600 MW(t). For electricity production, the MHR operates with an outlet helium temperature of 850 C to drive a direct, Brayton-cycle power-conversion system with a thermal-to-electrical conversion efficiency of 48 percent. This concept is referred to as the Gas Turbine MHR (GT-MHR). For hydrogen production, both electricity and process heat from the MHR are used to produce hydrogen. This concept is referred to as the H2-MHR. This report provides pre-conceptual design descriptions of full-scale, nth-of-a-kind H2 MHR plants based on thermochemical water splitting using the Sulfur-Iodine process and High-Temperature Electrolysis.

  4. Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II)

    E-Print Network [OSTI]

    Connaway, Heather M. (Heather Moira)

    2012-01-01

    The MIT Research Reactor (MITR-II) is currently undergoing analysis for the planned conversion from high enriched uranium (HEU) to low enriched uranium (LEU), as part of a global effort to minimize the availability of ...

  5. Antineutrino Monitoring of Thorium Reactors

    E-Print Network [OSTI]

    Akindele, Oluwatomi A; Norman, Eric B

    2015-01-01

    Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuels types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring.Thorium molten salt reactors (MSR) breed U-233, that if diverted constitute an IAEA direct use material. The antineutrino spectrum from the fission of U-233 has been determined, the feasibility of detecting the diversion of a significant quantity, 8 kg of U-233, within the IAEA timeliness goal of 30 days has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario at a 25 meter standoff by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos. It was found that the diversion of a signifi...

  6. 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 ...

  7. Granular Dynamics in Pebble Bed Reactor Cores

    E-Print Network [OSTI]

    Laufer, Michael Robert

    2013-01-01

    Design Criteria for Nucler Power Plants,” U.S. NuclearDesign Criteria GRECO Granular Recirculation Code HDPE High-Density Polyethylene HTR High-Temperature Reactor NRC Nuclear

  8. Evaluation of the feasibility and viability of modular pumped storage hydro (m-PSH) in the United States

    SciTech Connect (OSTI)

    Witt, Adam M.; Hadjerioua, Boualem; Martinez, Rocio; Bishop, Norm

    2015-09-01

    The viability of modular pumped storage hydro (m-PSH) is examined in detail through the conceptual design, cost scoping, and economic analysis of three case studies. Modular PSH refers to both the compactness of the project design and the proposed nature of product fabrication and performance. A modular project is assumed to consist of pre-fabricated standardized components and equipment, tested and assembled into modules before arrival on site. This technology strategy could enable m-PSH projects to deploy with less substantial civil construction and equipment component costs. The concept of m-PSH is technically feasible using currently available conventional pumping and turbine equipment, and may offer a path to reducing the project development cycle from inception to commissioning.

  9. Automatic safety rod for reactors

    DOE Patents [OSTI]

    Germer, John H. (San Jose, CA)

    1988-01-01

    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.

  10. The New Modular Control System for Power Converters at CERN

    E-Print Network [OSTI]

    Di Cosmo, Matteo

    2015-01-01

    The CERN accelerator complex consists of several generations of particle accelerators, with around 5000 power converters supplying regulated current and voltage to normal and superconducting magnet circuits. Today around 12 generations of converter control platforms can be found in the accelerator complex, ranging in age and technology. The diversity of these platforms has a significant impact on operability, maintenance and support of power converters. Over the past few years a new generation of modular controls called RegFGC3 has been developed by CERN’s power conversion group, with a goal to provide a standardised control platform, supporting a wide variety of converter topologies. The aim of this project is to reduce maintenance costs by decreasing the variety and diversity of control systems whilst simultaneously improving the operability and reliability of power converters and their controls. This paper describes the state of the on-going design and realization of the RegFGC3 platform, focusing on fun...

  11. Modular Energy Storage System for Hydrogen Fuel Cell Vehicles

    SciTech Connect (OSTI)

    Janice Thomas

    2010-05-31

    The objective of the project is to develop technologies, specifically power electronics, energy storage electronics and controls that provide efficient and effective energy management between electrically powered devices in alternative energy vehicles â?? plug-in electric vehicles, hybrid vehicles, range extended vehicles, and hydrogen-based fuel cell vehicles. The in-depth research into the complex interactions between the lower and higher voltage systems from data obtained via modeling, bench testing and instrumented vehicle data will allow an optimum system to be developed from a performance, cost, weight and size perspective. The subsystems are designed for modularity so that they may be used with different propulsion and energy delivery systems. This approach will allow expansion into new alternative energy vehicle markets.

  12. Design and Construction Solutions in the Accurate Realization of NCSX Magnetic Fields

    SciTech Connect (OSTI)

    Heitzenroeder, P.; Dudek, Lawrence E.; Brooks, Arthur W.; Viola, Michael E.; Brown, Thomas; Neilson, George H.; Zarnstorff, Michael C.; Rej, Donald; Cole,Michael J.; Freudenberg, Kevin D.; Harris J. H.; McGinnis, Gary

    2008-09-29

    The National Compact Stellarator Experiment, NCSX, is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge national Laboratory. The goal of NCSX is to provide the understanding necessary to develop an attractive, disruption free, steady state compact stellaratorbased reactor design. This paper describes the recently revised designs of the critical interfaces between the modular coils, the construction solutions developed to meet assembly tolerances, and the recently revised trim coil system that provides the required compensation to correct for the “as built” conditions and to allow flexibility in the disposition of as-built conditions. In May, 2008, the sponsor decided to terminate the NCSX project due to growth in the project’s cost and schedule estimates. However significant technical challenges in design and construction were overcome, greatly reducing the risk in the remaining work to complete the project.

  13. Modular multiplication operator and quantized baker's maps

    SciTech Connect (OSTI)

    Lakshminarayan, Arul [Max-Planck-Institut fuer Physik komplexer Systeme, Noethnitzer Strasse 38, D-01187 Dresden (Germany)

    2007-10-15

    The modular multiplication operator, a central subroutine in Shor's factoring algorithm, is shown to be a coherent superposition of two quantum baker's maps when the multiplier is 2. The classical limit of the maps being completely chaotic, it is shown that there exist perturbations that push the modular multiplication operator into regimes of generic quantum chaos with spectral fluctuations that are those of random matrices. For the initial state of relevance to Shor's algorithm we study fidelity decay due to phase and bit-flip errors in a single qubit and show exponential decay with shoulders at multiples or half-multiples of the order. A simple model is used to gain some understanding of this behavior.

  14. Copper vapor laser modular packaging assembly

    DOE Patents [OSTI]

    Alger, Terry W. (Tracy, CA); Ault, Earl R. (Dublin, CA); Moses, Edward I. (Castro Valley, CA)

    1992-01-01

    A modularized packaging arrangement for one or more copper vapor lasers and associated equipment is disclosed herein. This arrangement includes a single housing which contains the laser or lasers and all their associated equipment except power, water and neon, and means for bringing power, water, and neon which are necessary to the operation of the lasers into the container for use by the laser or lasers and their associated equipment.

  15. Copper vapor laser modular packaging assembly

    DOE Patents [OSTI]

    Alger, T.W.; Ault, E.R.; Moses, E.I.

    1992-12-01

    A modularized packaging arrangement for one or more copper vapor lasers and associated equipment is disclosed herein. This arrangement includes a single housing which contains the laser or lasers and all their associated equipment except power, water and neon, and means for bringing power, water, and neon which are necessary to the operation of the lasers into the container for use by the laser or lasers and their associated equipment. 2 figs.

  16. Modular architecture for robotics and teleoperation

    DOE Patents [OSTI]

    Anderson, Robert J. (11908 Ibex Ave., N.E., Albuquerque, NM 87111)

    1996-12-03

    Systems and methods for modularization and discretization of real-time robot, telerobot and teleoperation systems using passive, network based control laws. Modules consist of network one-ports and two-ports. Wave variables and position information are passed between modules. The behavior of each module is decomposed into uncoupled linear-time-invariant, and coupled, nonlinear memoryless elements and then are separately discretized.

  17. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

    1983-01-01

    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.

  18. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect (OSTI)

    R. Wigeland; J. Cahalan

    2009-09-01

    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.

  19. Modular Dynamical Semigroups for Quantum Dissipative Systems

    E-Print Network [OSTI]

    David Taj; Hans Christian Öttinger

    2015-03-10

    We introduce a class of Markovian quantum master equations, able to describe the dissipative dynamics of a quantum system weakly coupled to one or several heat baths. The dissipative structure is driven by an entropic operator, the so called modular Hamiltonian, which makes it nonlinear. The generated Modular Dynamical Semigroup (MDS) is not, in general, a Quantum Dynamical Semigroup (QDS), whose dynamics is of the popular Lindblad type. The MDS has a robust thermodynamic structure, which guarantees for the positivity of the time evolved state, the correct steady state properties, the positivity of the entropy production, a positive Onsager matrix and Onsager symmetry relations (arising from Green-Kubo formulas). We show that the celebrated Davies generator, obtained through the Born and the secular approximations, generates a MDS. By unravelling the modular structure of the former, we provide a different and genuinely nonlinear MDS, not of QDS type, which is free from the severe spectral restrictions of the Davies generator, while still being supported by a weak coupling limit argument. With respect to the latter, the present work is a substantial extension of \\cite{Ottinger2011_GEO,Ottinger2010_TLS_DHO}

  20. Engineering Development of Slurry Bubble Column Reactor (SBCR) Technology

    SciTech Connect (OSTI)

    Toseland, B.A.

    1998-10-29

    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.