WorldWideScience

Sample records for core design concepts

  1. Multimedia foundations core concepts for digital design

    CERN Document Server

    Costello, Vic; Youngblood, Susan

    2012-01-01

    Understand the core concepts and skills of multimedia production and digital storytelling using text, graphics, photographs, sound, motion, and video. Then, put it all together using the skills that you have developed for effective project planning, collaboration, visual communication, and graphic design. Presented in full color with hundreds of vibrant illustrations, Multimedia Foundations trains you in the principles and skill sets common to all forms of digital media production, enabling you to create successful, engaging content, no matter what tools you are using. Companion website

  2. Some concept for the TRIGA core design

    International Nuclear Information System (INIS)

    Aizawa, Otohiko

    1994-01-01

    There is the research reactor called TRIGA Mark-2 of 100 kW in Atomic Energy Research Laboratory, Musashi Institute of Technology. Recently, while the various calculations on the core were carried out, the author became aware of that this TRIGA core was designed at that time with excellent consideration. The reason for that is, although fuel is arranged in simple concentric circular state at a glance, it was known that in reality, this is the modification of the hexagonal core of triangular lattice. In the examination of square lattice fuel arrangement, the reactivity was calculated by using the gap between fuel rods as the parameter and by using ENDF/B-4 library and Monte Carlo code Keno-5. It is known that the design of the lattice with maximum reactivity cannot be done by the square lattice. The similar examination was carried out on triangular lattice, and it was found that the gap between fuel rods of 4 mm is the optimal design. The average neutron energy spectra in the fuel rods of the TRIGA Mark-2 core agreed considerably well with the energy spectra at 4.16 cm fuel rod pitch in triangular hexagonal core. In the reactor of about 100 kW, even if the gap between fuel rods is less than 4 mm, heat removal is sufficiently possible. (K.I.)

  3. Core design with respect to the safety concept

    International Nuclear Information System (INIS)

    Kollmar, W.

    1981-01-01

    In the present paper the following topics are dealt with: Principles of reactor core design and optimization, fuel management and safety concept for higher cycles and results of risk analyses (e.g. rod ejection, steam line break etc.) (RW)

  4. Soft shell hard core concept for aircraft impact resistant design

    International Nuclear Information System (INIS)

    Chen, C.; Rieck, P.J.

    1978-01-01

    For nuclear power plants sited in the vicinity of airports, the hypothetical events of aircraft impact have to be designed for. The conventional design concept is to strengthen the exterior structure to resist the impact induced force. The stiffened structures have two (2) disadvantages; one is the high construction cost, and the other is the high reaction force induced as well as the vibrational effects on the interior equipment and piping systems. This new soft shell hard core concept can relieve the above shortcomings. In this concept, the essential equipment required for safety are installed inside the hard core area for protection and the non-essential equipment are maintained between the hard core and soft shell area. During a hypothetical impact event, the soft shell will collapse locally and absorb large amounts of kinetic energy; hence, it reduces the reaction force and the vibrational effects. The design and analysis of the soft shell concept are discussed. (Author)

  5. Benefits of Low Boron Core Design Concept for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2009-10-15

    Nuclear design study was carried out to develop low boron core (LBC) based on one of current PWR concepts, OPR-1000. Most of design parameters were the same with those of Ulchin unit-5 except extensive utilization of burnable poison (BP) pins in order to compensate reactivity increase in LBC. For replacement of reduced soluble boron concentration, four different kinds of integral burnable absorbers (IBAs) such as gadolinia, integral fuel burnable absorber (IFBA), erbia and alumina boron carbide were considered in suppressing more excess reactivity. A parametric study was done to find the optimal core options from many design candidates for fuel assemblies and cores. Among them, the most feasible core design candidate was chosen in accordance with general design requirements. In this paper, the feasibility and design change benefits of the most favorable LBC design were investigated in more detail through the comparison of neutronic and thermal hydraulic design parameters of LBC with the reference plant (REF). As calculation tools, the HELIOS/MASTER code package and the MATRA code were utilized. The main purpose of research herein is to estimate feasibility and capability of LBC which was mainly designed to mitigate boron dilution accident (BDA), and for reduction of corrosion products. The LBC design concept using lower boron concentration with an elevated enrichment in {sup 10}B allows a reduction in the concentration of lithium in the primary coolant required to maintain the optimum coolant pH. All in all, LBC with operation at optimum pH is expected to achieve some benefits from radiation source reduction of reduced corrosion product, the limitation of the Axial Offset Anomaly (AOA) and fuel cladding corrosion. Additionally, several merits of LBC are closely related to fluid systems and system related aspects, reduced boron and lithium costs, equipment size reduction for boric acid systems, elimination of heat tracing, and more aggressive fuel design concepts.

  6. Benefits of Low Boron Core Design Concept for PWR

    International Nuclear Information System (INIS)

    Daing, Aung Tharn; Kim, Myung Hyun

    2009-01-01

    Nuclear design study was carried out to develop low boron core (LBC) based on one of current PWR concepts, OPR-1000. Most of design parameters were the same with those of Ulchin unit-5 except extensive utilization of burnable poison (BP) pins in order to compensate reactivity increase in LBC. For replacement of reduced soluble boron concentration, four different kinds of integral burnable absorbers (IBAs) such as gadolinia, integral fuel burnable absorber (IFBA), erbia and alumina boron carbide were considered in suppressing more excess reactivity. A parametric study was done to find the optimal core options from many design candidates for fuel assemblies and cores. Among them, the most feasible core design candidate was chosen in accordance with general design requirements. In this paper, the feasibility and design change benefits of the most favorable LBC design were investigated in more detail through the comparison of neutronic and thermal hydraulic design parameters of LBC with the reference plant (REF). As calculation tools, the HELIOS/MASTER code package and the MATRA code were utilized. The main purpose of research herein is to estimate feasibility and capability of LBC which was mainly designed to mitigate boron dilution accident (BDA), and for reduction of corrosion products. The LBC design concept using lower boron concentration with an elevated enrichment in 10 B allows a reduction in the concentration of lithium in the primary coolant required to maintain the optimum coolant pH. All in all, LBC with operation at optimum pH is expected to achieve some benefits from radiation source reduction of reduced corrosion product, the limitation of the Axial Offset Anomaly (AOA) and fuel cladding corrosion. Additionally, several merits of LBC are closely related to fluid systems and system related aspects, reduced boron and lithium costs, equipment size reduction for boric acid systems, elimination of heat tracing, and more aggressive fuel design concepts

  7. Core design concepts for high performance light water reactors

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.

    2007-01-01

    Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

  8. Core Thermal-Hydraulic Conceptual Design for the Advanced SFR Design Concepts

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Chang, Jin Wook; Yoo, Jae Woon; Song, Hoon; Choi, Sun Rock; Park, Won Seok; Kim, Sang Ji

    2010-01-01

    The Korea Atomic Energy Research Institute (KAERI) has developed the advanced SFR design concepts from 2007 to 2009 under the National longterm Nuclear R and D Program. Two types of core designs, 1,200 MWe breakeven and 600 MWe TRU burner core have been proposed and evaluated whether they meet the design requirements for the Gen IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. In generally, the core thermal hydraulic design is performed during the conceptual design phase to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power of each assembly because the conventional SFR core is composed of hundreds of ducted assemblies with hundreds of fuel rods. In carrying out the thermal and hydraulic design, special attention has to be paid to several performance parameters in order to assure proper performance and safety of fuel and core; the coolant boiling, fuel melting, structural integrity of the components, fuel-cladding eutectic melting, etc. The overall conceptual design procedure for core thermal and hydraulic conceptual design, i.e., flow grouping and peak pin temperature calculations, pressure drop calculations, steady-state and detailed sub-channel analysis is shown Figure 1. In the conceptual design phase, results of core thermal-hydraulic design for advanced design concepts, the core flow grouping, peak pin cladding mid-wall temperature, and pressure drop calculations, are summarized in this study

  9. Core Design Concept and Core Structural Material Development for a Prototype SFR

    International Nuclear Information System (INIS)

    Chang, Jinwook

    2013-01-01

    Core design Concept: – Initial core is Uranium metal fueled core, then it will evolve into TRU core; – Tight pressure drop constraint lowers power density; – Trade-off studies with relaxed pressure drop constraint (~0.4MPa) are on-going; – Major feature will be finalized this year. • KAERI is developing advanced cladding for high burnup fuel in Ptototype SFR: – Advanced cladding materials are now developing, which shows superior high temperature mechanical property to the conventional material; – Processing technologies related to tube making process are now developed to enhance high temperature mechanical propertyl – Preliminary HT9 cladding tube was manufactured and out-of pile mechanical properties were evaluated. Advanced cladding tube is now being developed and being prepared for irradiation test

  10. A multi-crucible core-catcher concept: Design considerations and basic results

    International Nuclear Information System (INIS)

    Szabo, I.

    1995-01-01

    A multi-crucible core-catcher concept to be implemented in new light water reactor containments has recently been proposed. This paper deals with conceptual design considerations and the various ways this type of core-catcher could be designed to meet requirements for reactor application. A systematic functional analysis of the multi-crucible core-catcher concept and the results of the preliminary design calculation are presented. Finally, the adequacy of the multi-crucible core-catcher concept for reactor application is discussed. (orig.)

  11. LMR design concepts for transuranic management in low sodium void worth cores

    International Nuclear Information System (INIS)

    Hill, R.N.

    1991-01-01

    The fuel cycle processing techniques and hard neutron spectrum of the integral Fast Reactor (IFR) metal fuel cycle have favorable characteristics for the management of transuranics; and the wide range of breeding characteristics available in metal fuelled cores provides for flexibility in transuranic management strategy. Previous studies indicate that most design options which decrease the breeding ratio also allow a decrease in sodium void worth; therefore, low void worths are achievable in transuranic burning (low breeding ratio) core designs. This paper describes numerous trade studies assessing various design options for a low void worth transuranic burner core. A flat annular core design appears to be a promising concept; the high leakage geometry yields a low breeding ratio and small sodium void worth. To allow flexibility in breeding characteristics, alternate design options which achieve fissile self-sufficiency are also evaluated. A self-sufficient core design which is interchangeable with the burner core and maintains a low sodium void worth is developed. (author)

  12. Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Werner, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, John C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, Robert C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dion, Axel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ananth, Krishnan P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-11-01

    The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave

  13. The KALIMER-600 Reactor Core Design Concept with Varying Fuel Cladding Thickness

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Jang, Jin Wook; Kim, Yeong Il

    2006-01-01

    Recently, Korea Atomic Energy Research Institute (KAERI) has developed a 600MWe sodium cooled fast reactor, the KALIMER-600 reactor core concept using single enrichment fuel. This reactor core concept is characterized by the following design targets : 1) Breakeven breeding (or fissile-self-sufficient) without any blanket, 2) Small burnup reactivity swing ( 23 n/cm 2 ). In the previous design, the single enrichment fuel concept was achieved by using the special fuel assembly designs where non-fuel rods (i.e., ZrH 1.8 , B 4 C, and dummy rods) were used. In particular, the moderator rods (ZrH 1.8 ) were used to reduce the sodium void worth and the fuel Doppler coefficient. But it has been known that this hydride moderator possesses relatively poor irradiation behavior at high temperature. In this paper, a new core design concept for use of single enrichment fuel is described. In this concept, the power flattening is achieved by using the core region wise cladding thicknesses but all non-fuel rods are removed to simplify the fuel assembly design

  14. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

  15. LMR design concepts for transuranic management in low sodium void worth cores

    International Nuclear Information System (INIS)

    Hill, R.N.

    1991-01-01

    The fuel cycle processing techniques and hard neuron spectrum of the Integral Fast Reactor (IFR) metal fuel cycle have favorable characteristics for the management of transuranics; and the wide range of breeding characteristics available in metal fuelled cores provides for flexibility in transuranic management strategy. Previous studies indicate that most design options which decrease the breeding ratio also show a decrease in sodium void worth; therefore, low void worths are achievable in transuranic burning (low breeding ratio) core designs. This paper describes numerous trade studies assessing various design options for a low void worth transuranic burner core. A flat annular core design appears to be a promising concept; the high leakage geometry yields a low breeding ratio and small sodium void worth. To allow flexibility in breeding characteristics, alternate design options which achieve fissile self-sufficiency are also evaluated. A self-sufficient core design which is interchangeable with the burner core and maintains a low sodium void worth is developed. 13 refs., 1 fig., 4 tabs

  16. Ex-vessel core catcher design requirements and preliminary concepts evaluation

    International Nuclear Information System (INIS)

    Friedland, A.J.; Tilbrook, R.W.

    1974-01-01

    As part of the overall study of the consequences of a hypothetical failure to scram following loss of pumping power, design requirements and preliminary concepts evaluation of an ex-vessel core catcher (EVCC) were performed. EVCC is the term applied to a class of devices whose primary objective is to provide a stable subcritical and coolable configuration within containment following a postulated accident in which it is assumed that core debris has penetrated the Reactor Vessel and Guard Vessel. Under these assumed conditions a set of functional requirements were developed for an EVCC and several concepts were evaluated. The studies were specifically directed toward the FFTF design considering the restraints imposed by the physical design and construction of the FFTF plant

  17. Small Launch Vehicle Design Approaches: Clustered Cores Compared with Multi-Stage Inline Concepts

    Science.gov (United States)

    Waters, Eric D.; Beers, Benjamin; Esther, Elizabeth; Philips, Alan; Threet, Grady E., Jr.

    2013-01-01

    In an effort to better define small launch vehicle design options two approaches were investigated from the small launch vehicle trade space. The primary focus was to evaluate a clustered common core design against a purpose built inline vehicle. Both designs focused on liquid oxygen (LOX) and rocket propellant grade kerosene (RP-1) stages with the terminal stage later evaluated as a LOX/methane (CH4) stage. A series of performance optimization runs were done in order to minimize gross liftoff weight (GLOW) including alternative thrust levels, delivery altitude for payload, vehicle length to diameter ratio, alternative engine feed systems, re-evaluation of mass growth allowances, passive versus active guidance systems, and rail and tower launch methods. Additionally manufacturability, cost, and operations also play a large role in the benefits and detriments for each design. Presented here is the Advanced Concepts Office's Earth to Orbit Launch Team methodology and high level discussion of the performance trades and trends of both small launch vehicle solutions along with design philosophies that shaped both concepts. Without putting forth a decree stating one approach is better than the other; this discussion is meant to educate the community at large and let the reader determine which architecture is truly the most economical; since each path has such a unique set of limitations and potential payoffs.

  18. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

  19. Some Core Contested Concepts

    Science.gov (United States)

    Chomsky, Noam

    2015-01-01

    Core concepts of language are highly contested. In some cases this is legitimate: real empirical and conceptual issues arise. In other cases, it seems that controversies are based on misunderstanding. A number of crucial cases are reviewed, and an approach to language is outlined that appears to have strong conceptual and empirical motivation, and…

  20. Some core contested concepts.

    Science.gov (United States)

    Chomsky, Noam

    2015-02-01

    Core concepts of language are highly contested. In some cases this is legitimate: real empirical and conceptual issues arise. In other cases, it seems that controversies are based on misunderstanding. A number of crucial cases are reviewed, and an approach to language is outlined that appears to have strong conceptual and empirical motivation, and to lead to conclusions about a number of significant issues that differ from some conventional beliefs.

  1. Radiotoxicity study of a boiling water reactor core design based on a thorium-uranium fuel concept

    International Nuclear Information System (INIS)

    Nunez C, A.; Espinosa P, G.

    2007-01-01

    Full text: The innovative design of a Boiling Water Reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to 233 U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR. A comparison of the toxicity of the spent fuel showed that toxicity is lower in the thorium cycle than other commercial fuels as UO 2 and MOX (uranium and plutonium) in case of the one-through cycle for LWR. (Author)

  2. Dynamic behavior of homogeneous and heterogeneous LMFBR core-design concepts

    International Nuclear Information System (INIS)

    Chang, Y.I.; Henryson, H. II; Orechwa, Y.; Su, S.F.; Greenman, G.; Blomquist, R.

    1981-01-01

    The emphasis is placed on obtaining an understanding of the inherent difference between homogeneous and heterogeneous core configurations regarding neutronic characteristics related to the dynamic behavior. The space-time neutronic and thermal-hydraulic behavior was analyzed in detail for various core configurations by using the FX2-TH, a two-dimensional kinetics code with thermal-hydraulic feedback. In addition, the relationship between the flux tilt and the fundamental-to-first harmonic eigenvalue separation, and the sodium void reactivity in heterogeneous cores were also sutdied

  3. Design of a boiling water reactor core based on an integrated blanket-seed thorium-uranium concept

    International Nuclear Information System (INIS)

    Nunez-Carrera, Alejandro; Francois, Juan Luis; Martin-del-Campo, Cecilia; Espinosa-Paredes, Gilberto

    2005-01-01

    This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket-seed (BS) assembly. The integrated BS 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. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the 233 U build-up. 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 235 U 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 fuel assembly

  4. Core mechanics and configuration behavior of advanced LMFBR core restraint concepts

    International Nuclear Information System (INIS)

    Fox, J.N.; Wei, B.C.

    1978-02-01

    Core restraint systems in LMFBRs maintain control of core mechanics and configuration behavior. Core restraint design is complex due to the close spacing between adjacent components, flux and temperature gradients, and irradiation-induced material property effects. Since the core assemblies interact with each other and transmit loads directly to the core restraint structural members, the core assemblies themselves are an integral part of the core restraint system. This paper presents an assessment of several advanced core restraint system and core assembly concepts relative to the expected performance of currently accepted designs. A recommended order for the development of the advanced concepts is also presented

  5. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  6. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong.

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

  7. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  8. Concept Design of a Gravity Core Cooling Tank as a Passive Residual Heat Removal System for a Research Reactor

    International Nuclear Information System (INIS)

    Lee, Kwonyeong; Chi, Daeyoung; Kim, Seong Hoon; Seo, Kyoungwoo; Yoon, Juhyeon

    2014-01-01

    A core downward flow is considered to use a plate type fuel because it is benefit to install the fuel in the core. If a flow inversion from a downward to upward flow in the core by a natural circulation is introduced within a high heat flux region of residual heat, the fuel fails instantly due to zero flow. Therefore, the core downward flow should be sufficiently maintained until the residual heat is in a low heat flux region. In a small power research reactor, inertia generated by a flywheel of the PCP can maintain a downward flow shortly and resolve the problem of a flow inversion. However, a high power research reactor more than 10 MW should have an additional method to have a longer downward flow until a low heat flux. Usually, other research reactors have selected an active residual heat removal system as a safety class. But, an active safety system is difficult to design and expensive to construct. A Gravity Core Cooling Tank (GCCT) beside the reactor pool with a Residual Heat Removal Pipe connecting two pools was developed and designed preliminarily as a passive residual heat removal system for an open-pool type research reactor. It is very simple to design and cheap to construct. Additionally, a non-safety, but active residual heat removal system is applied with the GCCT. It is a Pool Water Cooling and Purification System. It can improve the usability of the research reactor by removing the thermal waves, and purify the reactor pool, the Primary Cooling System, and the GCCT. Moreover, it can reduce the pool top radiation level

  9. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  10. KALIMER design concept report

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Kyu; Kim, Young Cheol; Kim, Young In; Kim, Young Gyun; Kim, Eui Kwang; Song, Hoon; Chung, Hyun Tai; Hwang, Woan; Nam, Cheol; Sub, Sim Yoon; Kim, Yeon Sik; Whan, Wim Myung; Min, Byung Tae; Yoo, Bong; Lee, Jae Han; Lee, Hyeong Yeon; Kim, Jong Bum; Koo, Gyeong Hoi; Ham, Chang Shik; Kwon, Kee Choon; Kim, Jung Taek; Park, Jae Chang; Lee, Jung Woon; Lee, Yong Hee; Kim, Chang Hwoi; Sim, Bong Shick; Hahn, Do Hee; Choi, Jong Hyeun; Kwon, Sang Woon

    1997-07-01

    KAERI is working for the development of KALIMER and work is being done for methodology development, experimental facility set up and design concept development. The development target of KALIMER has been set as to make KALIMER safer, more economic, more resistant to nuclear proliferation, and yield less impact on the environment. To achieve the target, study has been made for setting up the design concept of KALIMER including the assessment of various possible design alternatives. This report is the results of the study for the KALIMER concept study and describes the design concept of KALIMER. The developed design concept study and describes the design concept of KALIMER. The developed design concept is to be used as the starting point of the next development phase of conceptual design and the concept will be refined and modified in the conceptual design phase. The scope of the work has been set as the NSSS and essential BOP systems. For systems, NSSS and functionally related major BOP are covered. Sizing and specifying conceptual structure are covered for major equipment. Equipment and piping are arranged for the parts where the arrangement is critical in fulfilling the foresaid intention of setting up the KALIMER design concept. This report consists of 10 chapters. Chapter 2 is for the top level design requirements of KALIMER and it serves as the basis of KALIMER design concept development. Chapter 3 summarizes the KALIMER concept and describes the general design features. The remaining chapters are for specific systems. (author). 29 tabs., 37 figs.

  11. KALIMER design concept report

    International Nuclear Information System (INIS)

    Park, Chang Kyu; Kim, Young Cheol; Kim, Young In; Kim, Young Gyun; Kim, Eui Kwang; Song, Hoon; Chung, Hyun Tai; Hwang, Woan; Nam, Cheol; Sim Yoon Sub; Kim, Yeon Sik; Wim Myung Whan; Min, Byung Tae; Yoo, Bong; Lee, Jae Han; Lee, Hyeong Yeon; Kim, Jong Bum; Koo, Gyeong Hoi; Ham, Chang Shik; Kwon, Kee Choon; Kim, Jung Taek; Park, Jae Chang; Lee, Jung Woon; Lee, Yong Hee; Kim, Chang Hwoi; Sim, Bong Shick; Hahn, Do Hee; Choi, Jong Hyeun; Kwon, Sang Woon.

    1997-07-01

    KAERI is working for the development of KALIMER and work is being done for methodology development, experimental facility set up and design concept development. The development target of KALIMER has been set as to make KALIMER safer, more economic, more resistant to nuclear proliferation, and yield less impact on the environment. To achieve the target, study has been made for setting up the design concept of KALIMER including the assessment of various possible design alternatives. This report is the results of the study for the KALIMER concept study and describes the design concept of KALIMER. The developed design concept study and describes the design concept of KALIMER. The developed design concept is to be used as the starting point of the next development phase of conceptual design and the concept will be refined and modified in the conceptual design phase. The scope of the work has been set as the NSSS and essential BOP systems. For systems, NSSS and functionally related major BOP are covered. Sizing and specifying conceptual structure are covered for major equipment. Equipment and piping are arranged for the parts where the arrangement is critical in fulfilling the foresaid intention of setting up the KALIMER design concept. This report consists of 10 chapters. Chapter 2 is for the top level design requirements of KALIMER and it serves as the basis of KALIMER design concept development. Chapter 3 summarizes the KALIMER concept and describes the general design features. The remaining chapters are for specific systems. (author). 29 tabs., 37 figs

  12. LMFBR core design analysis

    International Nuclear Information System (INIS)

    Cho, M.; Yang, J.C.; Yoh, K.C.; Suk, S.D.; Soh, D.S.; Kim, Y.M.

    1980-01-01

    The design parameters of a commercial-scale fast breeder reactor which is currently under construction by regeneration of these data is preliminary analyzed. The analysis of nuclear and thermal characteristics as well as safety features of this reactor is emphasized. And the evaluation of the initial core mentioned in the system description is carried out in the areas of its kinetics and control system, and, at the same time, the flow distribution of sodium and temperature distribution of the initial FBR core system are calculated. (KAERI INIS Section)

  13. Fast reactor core concepts to improve transmutation efficiency

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

    2015-01-01

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate

  14. Statistical core design

    International Nuclear Information System (INIS)

    Oelkers, E.; Heller, A.S.; Farnsworth, D.A.; Kearfott, K.J.

    1978-01-01

    The report describes the statistical analysis of DNBR thermal-hydraulic margin of a 3800 MWt, 205-FA core under design overpower conditions. The analysis used LYNX-generated data at predetermined values of the input variables whose uncertainties were to be statistically combined. LYNX data were used to construct an efficient response surface model in the region of interest; the statistical analysis was accomplished through the evaluation of core reliability; utilizing propagation of the uncertainty distributions of the inputs. The response surface model was implemented in both the analytical error propagation and Monte Carlo Techniques. The basic structural units relating to the acceptance criteria are fuel pins. Therefore, the statistical population of pins with minimum DNBR values smaller than specified values is determined. The specified values are designated relative to the most probable and maximum design DNBR values on the power limiting pin used in present design analysis, so that gains over the present design criteria could be assessed for specified probabilistic acceptance criteria. The results are equivalent to gains ranging from 1.2 to 4.8 percent of rated power dependent on the acceptance criterion. The corresponding acceptance criteria range from 95 percent confidence that no pin will be in DNB to 99.9 percent of the pins, which are expected to avoid DNB

  15. Evaluation of a hanging core support concept for LMR application

    International Nuclear Information System (INIS)

    Burelbach, J.P.; Cha, B.K.; Huebotter, P.R.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.; Wu, T.S.

    1985-01-01

    The paper describes an innovative design concept for a liquid metal reactor (LMR) core support structure (CSS). A hanging core support structure is described and analyzed. The design offers inherent safety features, constructability advantages, and potential cost reductions. Some safety considerations are examined which include the in-service inspection (ISI), the backup support system and the structural behavior in a hypothetical case of a broken beam in the core support structure

  16. Automated Core Design

    International Nuclear Information System (INIS)

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2005-01-01

    Multistate searching methods are a subfield of distributed artificial intelligence that aims to provide both principles for construction of complex systems involving multiple states and mechanisms for coordination of independent agents' actions. This paper proposes a multistate searching algorithm with reinforcement learning for the automatic core design of a boiling water reactor. The characteristics of this algorithm are that the coupling structure and the coupling operation suitable for the assigned problem are assumed and an optimal solution is obtained by mutual interference in multistate transitions using multiagents. Calculations in an actual plant confirmed that the proposed algorithm increased the convergence ability of the optimization process

  17. The Core Journal Concept in Black Studies

    Science.gov (United States)

    Weissinger, Thomas

    2010-01-01

    Black Studies scholars have shown interest in the core journal concept. Indeed, the idea of core journals for the study of the Black experience has changed several times since 1940. While Black Studies scholars are citing Black Studies journals with frequency, they also cite traditional disciplinary journals a great deal of the time. However,…

  18. Core concept for long operating cycle simplified BWR (LSBWR)

    International Nuclear Information System (INIS)

    Kouji, Hiraiwa; Noriyuki, Yoshida; Mikihide, Nakamaru; Hideaki, Heki; Masanori, Aritomi

    2002-01-01

    An innovative core concept for a long operating cycle simplified BWR (LSBWR) is currently being developed under a Toshiba Corporation and Tokyo Institute of Technology joint study. In this core concept, the combination of enriched uranium oxide fuels and loose-pitched lattice is adopted for an easy application of natural circulation. A combination of enriched gadolinium and 0.7-times sized small bundle with peripheral-positioned gadolinium rod is also adopted as a key design concept for 15-year cycle operation. Based on three-dimensional nuclear and thermal hydraulic calculation, a nuclear design for fuel bundle has been determined. Core performance has been evaluated based on this bundle design and shows that thermal performance and reactivity characteristics meet core design criteria. Additionally, a control rod operation plan for an extension of control rod life has been successfully determined. (author)

  19. Axial heterogeneous core concept applied for super phoenix reactor

    International Nuclear Information System (INIS)

    Batista, J.L.; Renke, C.A.C.; Waintraub, M.; Santos Bastos, W. dos; Brito Aghina, L.O. de.

    1991-11-01

    Always maintaining the current design rules, this paper presents a parametric study on the type of axial heterogeneous core concept (CHA), utilizing a core of fast reactor Super Phenix type, reaching a maximum thermal burnup rate of 150000 M W d/t and being managed in single batch. (author)

  20. Development of high nickel austenitic steels for the application to fast reactor cores, (I). Alloy design with the aid of the d-electrons concept

    International Nuclear Information System (INIS)

    Murata, Yoshinori; Morinaga, Masahiko; Yukawa, Natsuo; Ukai, Shigeharu; Nomura, Shigeo; Okuda, Takanari; Harada, Makoto

    1999-01-01

    The design of high nickel austenitic steels for the core materials of the fast reactors was performed following the d-electrons concept devised on the basis of molecular orbital calculations of transition-metal based alloys. In this design two calculated parameters are mainly utilized. The one is the d-orbital energy level (Md) of alloying transition elements, and the other is the bond order (Bo) that is a measure of the covalent bond strength between atoms. Using the Md-bar - Bo-bar phase stability diagram accurate prediction become possible for the phase stability of the austenite phase and 5% swelling at 140 dpa for nickel ions. Here, Md-bar and Bo-bar are the compositional average of Md and Bo parameters, respectively. On the basis of the phase stability diagram and preliminary experiments, guidelines for the alloy design of carbo-nitrides precipitated high nickel austenitic steels were constructed. Following the guidelines several new austenitic steels were designed for the fast reactors core material. (author)

  1. Neutronic studies of the long life core concept: Part 1, Design and performance of 1000 MWe uranium oxide fueled low power density LMR cores

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1987-04-01

    The parametric behavior of some key neutronic performance parameters for low power density LMR cores fueled with uranium oxide is investigated. The results are compared to reference homogeneous and heterogeneous cores with normal fuel management and Pu fueling. It can be concluded that with respect to minimizing the initial fissile mass and thereby economizing on the inventory costs and carrying charges, the superior neutron economy of the LMR fuel cycle is best exploited through normal fuel management with Pu recycling. In the once-through mode the LMR fuel cycle has disadvantages due to a higher fissile inventory and is not competitive with the LWR fuel cycle

  2. Validation of a Computational Model for the SLS Core Stage Oxygen Tank Diffuser Concept and the Low Profile Diffuser - An Advanced Development Design for the SLS

    Science.gov (United States)

    Brodnick, Jacob; Richardson, Brian; Ramachandran, Narayanan

    2015-01-01

    The Low Profile Diffuser (LPD) project originated as an award from the Marshall Space Flight Center (MSFC) Advanced Development (ADO) office to the Main Propulsion Systems Branch (ER22). The task was created to develop and test an LPD concept that could produce comparable performance to a larger, traditionally designed, ullage gas diffuser while occupying a smaller volume envelope. Historically, ullage gas diffusers have been large, bulky devices that occupy a significant portion of the propellant tank, decreasing the tank volume available for propellant. Ullage pressurization of spacecraft propellant tanks is required to prevent boil-off of cryogenic propellants and to provide a positive pressure for propellant extraction. To achieve this, ullage gas diffusers must slow hot, high-pressure gas entering a propellant tank from supersonic speeds to only a few meters per second. Decreasing the incoming gas velocity is typically accomplished through expansion to larger areas within the diffuser which has traditionally led to large diffuser lengths. The Fluid Dynamics Branch (ER42) developed and applied advanced Computational Fluid Dynamics (CFD) analysis methods in order to mature the LPD design from and initial concept to an optimized test prototype and to provide extremely accurate pre-test predictions of diffuser performance. Additionally, the diffuser concept for the Core Stage of the Space Launch System (SLS) was analyzed in a short amount of time to guide test data collection efforts of the qualification of the device. CFD analysis of the SLS diffuser design provided new insights into the functioning of the device and was qualitatively validated against hot wire anemometry of the exterior flow field. Rigorous data analysis of the measurements was performed on static and dynamic pressure data, data from two microphones, accelerometers and hot wire anemometry with automated traverse. Feasibility of the LPD concept and validation of the computational model were

  3. Development of core design technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; In, Kim Young; Kim, Young Il; Kim, Y G; Kim, S J; Song, H; Kim, T K; Kim, W S; Hwang, W; Lee, B O; Park, C K; Joo, H K; Yoo, J W; Kang, H Y; Park, W S

    2000-05-01

    For the development of KALIMER (150 MWe) core conceptual design, design evolution and optimization for improved economics and safety enhancement was performed in the uranium metallic fueled equilibrium core design which uses U-Zr binary fuel not in excess of 20 percent enrichment. Utilizing results of the uranium ,metallic fueled core design, the breeder equilibrium core design with breeding ratio being over 1.1 was developed. In addition, utilizing LMR's excellent neutron economy, various core concepts for minor actinide burnup, inherent safety, economics and non-proliferation were realized and its optimization studies were performed. A code system for the LMR core conceptual design has been established through the implementation of needed functions into the existing codes and development of codes. To improve the accuracy of the core design, a multi-dimensional nodal transport code SOLTRAN, a three-dimensional transient code analysis code STEP, MATRA-LMR and ASSY-P for T/H analysis are under development. Through the automation of design calculations for efficient core design, an input generator and several interface codes have been developed. (author)

  4. Development of core design technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Kim Young In; Kim, Young Il; Kim, Y. G.; Kim, S. J.; Song, H.; Kim, T. K.; Kim, W. S.; Hwang, W.; Lee, B. O.; Park, C. K.; Joo, H. K.; Yoo, J. W.; Kang, H. Y.; Park, W. S

    2000-05-01

    For the development of KALIMER (150 MWe) core conceptual design, design evolution and optimization for improved economics and safety enhancement was performed in the uranium metallic fueled equilibrium core design which uses U-Zr binary fuel not in excess of 20 percent enrichment. Utilizing results of the uranium ,metallic fueled core design, the breeder equilibrium core design with breeding ratio being over 1.1 was developed. In addition, utilizing LMR's excellent neutron economy, various core concepts for minor actinide burnup, inherent safety, economics and non-proliferation were realized and its optimization studies were performed. A code system for the LMR core conceptual design has been established through the implementation of needed functions into the existing codes and development of codes. To improve the accuracy of the core design, a multi-dimensional nodal transport code SOLTRAN, a three-dimensional transient code analysis code STEP, MATRA-LMR and ASSY-P for T/H analysis are under development. Through the automation of design calculations for efficient core design, an input generator and several interface codes have been developed. (author)

  5. The Seismographic Design Concept

    DEFF Research Database (Denmark)

    Salamon, Karen Lisa; Engholm, Ida

    2015-01-01

    This article gives an overview of the theoretical development of the design concept through two centuries in Europe and North America. Drawing on the academic disciplines of design history and anthropology, the authors present seminal moments in the theorization of “design”. Historically formativ...... argues for a more historically reflective glance on theory’s influence on the moulding of practice from ideology also in the context of design, and presents itself as a step in this self reflective direction....

  6. Design study on metal fuel FBR cores

    International Nuclear Information System (INIS)

    Yokoo, T.; Tanaka, Y.; Ogata, T.

    1991-01-01

    A design approach for metal fuel FBR core to maintain fuel integrity during transient events by limiting eutectic/liquid phase formation is proposed based on the current status of metallic fuel development. Its impact as the limitation on the core outlet temperature is assessed through its application to two of CRIEPI's core concepts, high linear power 1000 MWe homogeneous design and medium linear power 300 MWe radially heterogeneous design. SESAME/SALT code is used in this study to analyze steady state and transient fuel behavior. SE2-FA code is developed based on SUPERENERGY-2 and used to analyze core thermal-hydraulics with uncertainties. As the result, the core outlet temperatures of both designs are found to be limited to ≤500degC if it is required to prevent eutectic/liquid phase formation during operational transients in order to guarantee the fuel integrity. Additional assessment is made assuming an advanced limiting condition that allows small liquid phase formation based on the liquid phase penetration rate derived from existing experimental results. The result indicates possibility of raising core outlet temperature to ∼ 530degC. Also, it is found that core design technology improvements such as hot spot factors reduction can contribute to the core outlet temperature extension by 10 ∼ 20degC. (author)

  7. Innovative reactor core: potentialities and design

    International Nuclear Information System (INIS)

    Artioli, C.; Petrovich, Carlo; Grasso, Giacomo

    2010-01-01

    Gen IV nuclear reactors are considered a very attractive answer for the demand of energy. Because public acceptance they have to fulfil very clearly the requirement of sustainable development. In this sense a reactor concept, having by itself a rather no significant interaction with the environment both on the front and back end ('adiabatic concept'), is vital. This goal in mind, a new way of designing such a core has to be assumed. The starting point must be the 'zero impact'. Therefore the core will be designed having as basic constraints: a) fed with only natural or depleted Uranium, and b) discharges only fission products. Meantime its potentiality as a net burner of Minor Actinide has to be carefully estimated. This activity, referred to the ELSY reactor, shows how to design such an 'adiabatic' core and states its reasonable capability of burning MA legacy in the order of 25-50 kg/GW e y. (authors)

  8. Core catcher concepts future PWR-Plants

    International Nuclear Information System (INIS)

    Alsmeyer, H.; Werle, H.

    1994-01-01

    Light water reactors of the next generation should have still greater passive safety, even in the most serious accidents. This includes the long term safe inclusion of the core inventory in the case of core meltdown accidents. The three concepts for cooling the liquefied core outside the reactor pressure vessel examined by KfK should remove the post-shutdown heat by direct contact of the melt with water. The geometric distribution of the melt increases its surface area, so that favourable conditions for heat removal from the poorly thermally-conducting melt are created and complete quick solidification occurs. The experiments examine both the relocation and distribution mechanisms of the melt and the reactions occurring when water enters. As strong interaction is possible on direct contact of the melt with water, an important aim is experimental determination and limitation of any resulting mechanical stresses. (orig./HP) [de

  9. Toward full MOX core design

    International Nuclear Information System (INIS)

    Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

    1999-01-01

    This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

  10. Sellafield repository design concept

    International Nuclear Information System (INIS)

    1998-01-01

    Between 1989 and 1997, UK Nirex Ltd carried out a programme of investigations to evaluate the potential of a site adjacent to the BNFL Sellafield works to host a deep repository for the United Kingdom's intermediate-level and certain low-level radioactive waste. The programme of investigations was wound down following the decision in March 1997 to uphold the rejection of the Company's planning application for the Rock Characterisation Facility (RCF), an underground laboratory which would have allowed further investigations to confirm whether or not the site would be suitable. Since that time, the Company's efforts in relation to the Sellafield site have been directed towards documenting and publishing the work carried out. The design concept for a repository at Sellafield was developed in parallel with the site investigations through an iterative process as knowledge of the site and understanding of the repository system performance increased. This report documents the Sellafield repository design concept as it had been developed, from initial design considerations in 1991 up to the point when the RCF planning application was rejected. It shows, from the context of a project at that particular site, how much information and experience has been gained that will be applicable to the development of a deep waste repository at other potential sites

  11. A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

    International Nuclear Information System (INIS)

    Bae, Gonghoon; Hong, Ser Gi

    2015-01-01

    In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO 2 –ThO 2 fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO 2 –ThO 2 fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO 2 –UO 2 fuel pins are employed to achieve long-cycle length of ∼4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods. (author)

  12. Reliable core thermal design

    International Nuclear Information System (INIS)

    Amendola, A.

    1974-01-01

    The hot spot analysis is no longer limited to the calculation of a simple safety factor against overtemperature, but is now integrated in the overall design philosophy. This paper describes the development of a probabilistic method of analysis and compares it with other advanced calculation methods. Feedbacks from the analysis act: - on the nominal temperature distribution in order to satisfy the maximum temperature limit and in the same time to optimize the coolant temperature for maximum plant efficiency, and - on the specifications of manufacturing tolerances and experimental investigations in order to identify and to reduce the most important design uncertainties. Moreover the computer codes SHOSPA and THEDRA are briefly discussed. Both codes deliver the zero hot spot probability as a function of the geometrical size assumed for a ''spot''. THEDRA delivers also the expected hot spot distribution. By means of THEDRA it is possible to evaluate the pins failure expectation if the distribution of pin failures versus operating temperature is known. (author)

  13. Development of core design and analyses technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  14. Development of core design and analyses technology for integral reactor

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  15. LMFBR design and its evolution. (2) Core design of LMFBR

    International Nuclear Information System (INIS)

    Uto, Nariaki; Mizuno, Tomoyasu

    2003-01-01

    Sodium-cooled core design studies are performed. MOX fuel core with axial blanket partial elimination subassembly due to safety consideration is studied. This type of core with high internal conversion ratio possesses capability of achieving 26 months of operation cycle length and 100 GWd/t of burnup averaged over core and blanket, which are superior characteristics in view of reducing cost of power generation. Metal fuel core is also studied, and its higher breeding capability reveals a potential of better core performance such as longer operation cycle length for the same level of electricity generation, though core outlet temperature is limited to lower level due to steel cladding-metal fuel compatibility concerns. Another metal fuel core concept using single Pu enrichment and two radial regions with individual fuel pin diameters achieves 550degC of core outlet temperature identical to that of MOX fuel core, keeping operation cycle length comparable with that of MOX fuel core. This series of study results show that sodium-cooled MOX and metal fuel cores have a high flexibility in satisfying various needs including fuel cycle cost and breeding capability, depending on the stage of introducing commercialized fast reactor cycle system. (author)

  16. Borehole disposal design concept

    International Nuclear Information System (INIS)

    RANDRIAMAROLAHY, J.N.

    2007-01-01

    In Madagascar, the sealed radioactive sources are used in several socioeconomic sectors such as medicine, industry, research and agriculture. At the end of their useful lives, these radioactive sources become radioactive waste and can be still dangerous because they can cause harmful effects to the public and the environment. This work entitled 'Borehole disposal design concept' consists in putting in place a site of sure storage of the radioactive waste, in particular, sealed radioactive sources. Several technical aspects must be respected to carry out such a site like the geological, geomorphologic, hydrogeologic, geochemical, meteorological and demographic conditions. This type of storage is favorable for the developing countries because it is technologically simple and economic. The cost of construction depends on the volume of waste to store and the depth of the Borehole. The Borehole disposal concept provides a good level of safety to avoid the human intrusion. The future protection of the generations against the propagation of the radiations ionizing is then assured. [fr

  17. Development of Core Design Technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Il; Hong, S. G.; Jang, J. W. (and others)

    2007-06-15

    This report describes the contents of core design technology and computer code system development performed during 2005 and 2006 on the objects of nuclear proliferation resistant core and nuclear fuel basic key technology development security. Also, it is including the future application plans for the results and the developed methodology, important information and the materials acquired in this period. Two core designs with single enrichment were considered for the KALIMER-600 during the first year : 1) the first core uses the non-fuel rods such as B4C, ZrH1.8, and dummy rods, 2) the core using different cladding thickness for each core region (inner, middle, and outer cores) without non-fuel rods to flatten the power distribution. In particular, the latter design was intended to simplify the fuel assembly design by eliminating the heterogeneity. It was found that the proposed design satisfy all of the Gen IV SFR design goals on the cycle length longer than 18 EFPM, fuel discharge burnup larger than 80GWd/t, sodium void worth, conversion ratio, reactivity burnup swing and so on. For this object reactor, the structure integrity outside of reactor is confirmed for the radiation exposure during the plant life according to the result of shielding design and evaluation. The transmutation capability and the core characteristics of sodium cooled fast reactor was also evaluated according to the change of MA amount. The reactivity coefficients for the BN-600 reactor with MA fueled are calculated and the results are compared and evaluated with other participants results. Even though the discrepancies between the results of participants are somewhat large but the K-CORE results are close to the average within a standard deviation. To have the capability of 3-dimensional core dynamic analysis such as analyzing power distribution and reactivity variations according to the asymmetric insertion/withdrawal of control rods, the calculation module for core dynamic parameters was

  18. A core design study for 'zero-sodium-void-worth' cores

    International Nuclear Information System (INIS)

    Kawashima, Masatoshi; Suzuki, Masao; Hill, R.N.

    1992-01-01

    Recently, a number of low sodium-void-worth metal-fueled core design concepts have been proposed; to provide for flexibility in transuranic nuclide management strategy, core designs which exhibit a wide range of breeding characteristics have been developed. Two core concepts, a flat annular (transuranic burning) core and an absorber-type parfait (transuranic self-sufficient) core, are selected for this study. In this paper, the excess reactivity management schemes applied in the two designs are investigated in detail. In addition, the transient effect of reactivity insertions on the parfait core design is assessed. The upper and lower core regions in the parfait design are neutronically decoupled; however, the common coolant channel creates thermalhydraulic coupling. This combination of neutronic and thermalhydraulic characteristics leads to unique behavior in anticipated transient overpower events. (author)

  19. SMART core protection system design

    International Nuclear Information System (INIS)

    Lee, J. K.; Park, H. Y.; Koo, I. S.; Park, H. S.; Kim, J. S.; Son, C. H.

    2003-01-01

    SMART COre Protection System(SCOPS) is designed with real-tims Digital Signal Processor(DSP) board and Network Interface Card(NIC) board. SCOPS has a Control Rod POSition (CRPOS) software module while Core Protection Calculator System(CPCS) consists of Core Protection Calculators(CPCs) and Control Element Assembly(CEA) Calculators(CEACs) in the commercial nuclear plant. It's not necessary to have a independent cabinets for SCOPS because SCOPS is physically very small. Then SCOPS is designed to share the cabinets with Plant Protection System(PPS) of SMART. Therefor it's very easy to maintain the system because CRPOS module is used instead of the computer with operating system

  20. Conceptual Models Core to Good Design

    CERN Document Server

    Johnson, Jeff

    2011-01-01

    People make use of software applications in their activities, applying them as tools in carrying out tasks. That this use should be good for people--easy, effective, efficient, and enjoyable--is a principal goal of design. In this book, we present the notion of Conceptual Models, and argue that Conceptual Models are core to achieving good design. From years of helping companies create software applications, we have come to believe that building applications without Conceptual Models is just asking for designs that will be confusing and difficult to learn, remember, and use. We show how Concept

  1. An axially and radially two-zoned large liquid-metal fast breeder reactor core concept

    International Nuclear Information System (INIS)

    Kamei, T.; Arie, K.; Moriki, Y.; Suzuki, M.; Yamaoka, M.

    1985-01-01

    A new core concept that has advantages over conventional homogeneous cores in neutronics characteristics such as power peaking factor, burnup reactivity loss, and reactivity response to the movement of control rods in earthquakes has been evolved. Two options of the new core concept are feasible. One is the so-called axially heterogeneous core, with the internal blanket placed at the lower part of the core. The other concept is similar to the conventional homogeneous core, but has two different plutonium-enriched zones in the axial as well as in the radial direction, so it is a hybrid type of the conventional homogeneous core and the axially heterogeneous core. The new design concept is described and the way that the core characteristics are improved by the chosen key parameters is shown

  2. Designing concepts and strategies

    DEFF Research Database (Denmark)

    Kiib, Hans

    2012-01-01

    , that new developments often employ very modest research on the subject and often very little has been done in order to challenge traditional concepts and to invent new sustainable concepts for redevelopment. In order to avoid mistakes in urban redevelopment we need to learn from research and evaluation...... of the best planning practice. But what might be just as important is to learn from concept development practice, which can give us a comprehensive understanding of our complex cities and make us develop a way of experiencing the unique qualities of the architectural typologies at the site. Finally...... and strategies are briefly described in the article, and the adaption by city planners and developers has been critical reviewed....

  3. Conceptual design of PFBR core

    International Nuclear Information System (INIS)

    Lee, S.M.; Govindarajan, S.; Indira, R.; John, T.M.; Mohanakrishnan, P.; Shankar Singh, R.; Bhoje, S.B.

    1996-01-01

    The design options selected for the core of the 500 MWe Prototype Fast Breeder Reactor are presented. PFBR has a conventional mixed oxide fuel core of homogeneous type with two enrichment zones for power flattening and with radial and axial blankets to make the reactor self-sustaining in fissile material. Pin diameter has been selected for minimization of fissile inventory. Considerations for the choice of number of pins per subassembly, integrated versus separate axial blankets, and other pin and subassembly parameters are discussed. As the core size is moderate, no special schemes for reducing the maximum positive sodium voiding coefficient is envisages. Two independent, diverse fast acting shutdown systems working in fail-safe mode are selected. The number of absorber rods has been minimized by choosing a layout for maximum antishadow effect. Nine control and safety rods are distributed in two rods for power flattening by differential insertion. Three Diverse Safety Rods, are also provided which are normally fully withdrawn. The optimization of layout of radial and axial shielding and adequacy of flux at detector location are also discussed. (author). 2 figs

  4. Review of advanced core designs for LMFBRs

    International Nuclear Information System (INIS)

    Yoshida, Kazuo

    1986-01-01

    It is a matter of great importance for the development of LMFBR to reduce its power cost to the level of the other power generating means. For this purpose, some ideas that use advanced core concepts to reduce LMFBR's power cost by improving its fuel cycle economics have recently been proposed. In this report, two hopeful ideas that use advanced core concepts: (1) Ultra Long Life Core (ULLC) - non-refueling over LMFBR power plant life; (2) Integral Fast Reactor (IFR) concept - metal fueled core and pyrometallurgical reprocessing; are picked up and their economical effect and technical probrems are investigated. (author)

  5. Progress of full MOX core design in ABWR

    International Nuclear Information System (INIS)

    Izutsu, S.; Sasagawa, M.; Aoyama, M.; Maruyama, H.; Suzuki, T.

    2000-01-01

    Full MOX ABWR core design has been made, based on the MOX design concept of 8x8 bundle configuration with a large central water rod, 40 GWd/t maximum bundle exposure, and the compatibility with 9x9 high-burnup UO 2 bundles. Core performance on shutdown margin and thermal margin of the MOX-loaded core is similar to that of UO 2 cores for the range from full UO 2 core to full MOX core. Safety analyses based on its safety parameters and MOX property have shown its conformity to the design criteria in Japan. In order to confirm the applicability of the nuclear design method to full MOX cores, Tank-type Critical Assembly (TCA) experiment data have been analyzed on criticality, power distribution and β eff /l measurements. (author)

  6. Identifying Core Concepts of Cybersecurity: Results of Two Delphi Processes

    Science.gov (United States)

    Parekh, Geet; DeLatte, David; Herman, Geoffrey L.; Oliva, Linda; Phatak, Dhananjay; Scheponik, Travis; Sherman, Alan T.

    2018-01-01

    This paper presents and analyzes results of two Delphi processes that polled cybersecurity experts to rate cybersecurity topics based on importance, difficulty, and timelessness. These ratings can be used to identify core concepts--cross-cutting ideas that connect knowledge in the discipline. The first Delphi process identified core concepts that…

  7. Design configuration of GCFR core assemblies

    International Nuclear Information System (INIS)

    LaBar, M.P.; Lee, G.E.; Meyer, R.J.

    1980-05-01

    The current design configurations of the core assemblies for the gas-cooled fast reactor (GCFR) demonstration plant reactor core conceptual design are described. Primary emphasis is placed upon the design innovations that have been incorporated in the design of the core assemblies since the establishment of the initial design of an upflow GCFR core. A major feature of the design configurations is that they are prototypical of core assemblies for use in commercial plants; a larger number of the same assemblies would be used in a commercial plant

  8. New approach to the design of core support structures for large LMFBR plants

    International Nuclear Information System (INIS)

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-01-01

    The paper describes an innovative design concept for a LMFBR Core Support Structure. A hanging Core Support Structure is described and analyzed. The design offers inherent safety features, constructibility advantages, and potential cost reductions

  9. Design Principles for Synthesizable Processor Cores

    DEFF Research Database (Denmark)

    Schleuniger, Pascal; McKee, Sally A.; Karlsson, Sven

    2012-01-01

    As FPGAs get more competitive, synthesizable processor cores become an attractive choice for embedded computing. Currently popular commercial processor cores do not fully exploit current FPGA architectures. In this paper, we propose general design principles to increase instruction throughput...

  10. ETHICAL FASHION CONCEPT AND DESIGNERS

    Directory of Open Access Journals (Sweden)

    Pinar GOKLUBERK OZLU

    2015-01-01

    Full Text Available Some problems like rapidly developing industrialization, irregular population growth, environmental pollution and to feel the impact of global warming as seriously, has been giving significant damage to the earth. People has realized that, after polluting to clean is harder than polluting of the measures to be taken before. And again people showed the sensitivity to the environment through different reactions and sanctions, took measures and created the new concepts about the enviroment. "Ethical Fashion" concept was created by the conscious and responsible individuals in the last two decades. However, that are being implemented as a concept is noticeable. Textile and fashion industry cover "Ethical Fashion"; ecological product, working conditions, fair trade and sustainable product are all in that concept. "Ethical Fashion" appeared and developed especially in United Kingdom, the USA and the other European countries. Nowadays, we may see a lot of textile and fashion designers, fabric and clothing collections, fairs and some specific courses at the universities about "Ethical Fashion". In this research contains "Ethical Fashion" concept, it's development processes and fashion designers who is working for this concept at the present time, also the main target is in this research, semtinizing "Ethical Fashion" concept.

  11. Westinghouse Nuclear Core Design Training Center - a design simulator

    International Nuclear Information System (INIS)

    Altomare, S.; Pritchett, J.; Altman, D.

    1992-01-01

    The emergence of more powerful computing technology enables nuclear design calculations to be done on workstations. This shift to workstation usage has already had a profound effect in the training area. In 1991, the Westinghouse Electric Corporation's Commercial Nuclear Fuel Division (CNFD) developed and implemented a Nuclear Core Design Training Center (CDTC), a new concept in on-the-job training. The CDTC provides controlled on-the-job training in a structured classroom environment. It alllows one trainer, with the use of a specially prepared training facility, to provide full-scope, hands-on training to many trainees at one time. Also, the CDTC system reduces the overall cycle time required to complete the total training experience while also providing the flexibility of individual training in selected modules of interest. This paper provides descriptions of the CDTC and the respective experience gained in the application of this new concept

  12. Core concepts of spatial prioritisation in systematic conservation planning.

    Science.gov (United States)

    Kukkala, Aija S; Moilanen, Atte

    2013-05-01

    Systematic conservation planning (SCP) is a field of conservation biology concerned with delivering on-the-ground actions that achieve conservation goals. It describes a set of operational models that cover both design and implementation of conservation, with a strong focus on mobilising the collective action typically required to implement conservation. SCP, as it was originally described, was composed of six different stages: collection of data, identification of conservation goals, evaluation of the existing protected area network, design of expansions, implementation of conservation action, and long-term maintenance of biodiversity in the network. Since then, the operational model has been expanded into several different variants. Conservation actions applied inside SCP include establishment and expansion of reserve networks and allocation of habitat restoration and management. Within the broader context of SCP, there is a fundamental biogeographic-economic analysis frequently called spatial conservation prioritisation or conservation assessment, which is used for identifying where important areas for biodiversity are and how conservation goals might be achieved efficiently. Here, we review the usage and meaning of the 12 biogeographic-economic core concepts of SCP: adequacy, complementarity, comprehensiveness, effectiveness, efficiency, flexibility, irreplaceability, replacement cost, representation, representativeness, threat, and vulnerability. Some of the concepts have clear definitions whereas others may have alternative and possibly conflicting definitions. With a comprehensive literature review literature, we elucidate the historical backgrounds of these concepts, the first definitions and usages, alternative later definitions, key applications, and prior reviews. This review reduces linguistic uncertainty in the application of SCP. Since SCP is a global activity with a multitude of different stakeholders involved, it is vital that those involved can

  13. Core melt retention and cooling concept of the ERP

    Energy Technology Data Exchange (ETDEWEB)

    Weisshaeupl, H [SIEMENS/KWU, Erlangen (Germany); Yvon, M [Nuclear Power International, Paris (France)

    1996-12-01

    For the French/German European Pressurized Water Reactor (EPR) mitigative measures to cope with the event of a severe accident with core melt down are considered already at the design stage. Following the course of a postulated severe accident with reactor pressure vessel melt through one of the most important features of a future design must be to stabilize and cool the melt within the containment by dedicated measures. This measures should - as far as possible - be passive. One very promising solution for core melt retention seems to be a large enough spreading of the melt on a high temperature resistant protection layer with water cooling from above. This is the favorite concept for the EPR. In dealing with the retention of a molten core outside of the RPV several ``steps`` from leaving the RPV to finally stabilize the melt have to gone through. These steps are: collection of the melt; transfer of the melt; distribution of the melt; confining; cooling and stabilization. The technical features for the EPR solution of a large spreading of the melt are: Dedicated spreading chamber outside the reactor pit (area about 150 m{sup 2}); high temperature resistant protection layers (e.g. Zirconia bricks) at the bottom and part of the lateral structures (thus avoiding melt concrete interaction); reactor pit and spreading compartment are connected via a discharge channel which has a slope to the spreading area and is closed by a steel plate, which will resist the core melt for a certain time in order to allow a collection of the melt; the spreading compartments is connected with the In-Containment Refuelling Water Storage Tank (IRWST) with pipes for water flooding after spreading. These pipes are closed and will only be opened by the hot melt itself. It is shown how the course of the different steps mentioned above is processed and how each of these steps is automatically and passively achieved. (Abstract Truncated)

  14. Small ex-core heat pipe thermionic reactor concept (SEHPTR)

    International Nuclear Information System (INIS)

    Jacox, M.G.; Bennett, R.G.; Lundberg, L.B.; Miller, B.G.; Drexler, R.L.

    1991-01-01

    The Idaho National Engineering Laboratory (INEL) has developed an innovative space nuclear power concept with unique features and significant advantages for both Defense and Civilian space missions. The Small Ex-core Heat Pipe Thermionic Reactor (SEHPTR) concept was developed in response to Air Force needs for space nuclear power in the range of 10 to 40 kilowatts. This paper describes the SEHPTR concept and discusses the key technical issues and advantages of such a system

  15. Development of core design and analyses technology for integral reactor

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Lee, C. C.; Kim, K. Y.

    2002-03-01

    In general, small and medium-sized integral reactors adopt new technology such as passive and inherent safety concepts to minimize the necessity of power source and operator actions, and to provide the automatic measures to cope with any accidents. Specifically, such reactors are often designed with a lower core power density and with soluble boron free concept for system simplification. Those reactors require ultra long cycle operation for higher economical efficiency. This cycle length requirement is one of the important factors in the design of burnable absorbers as well as assurance of shutdown margin. Hence, both computer code system and design methodology based on the today's design technology for the current commercial reactor cores require intensive improvement for the small and medium-sized soluble boron free reactors. New database is also required for the development of this type of reactor core. Under these technical requirements, conceptual design of small integral reactor SMART has been performed since July 1997, and recently completed under the long term nuclear R and D program. Thus, the final objectives of this work is design and development of an integral reactor core and development of necessary indigenous design technology. To reach the goal of the 2nd stage R and D program for basic design of SMART, design bases and requirements adequate for ultra long cycle and soluble boron free concept are established. These bases and requirements are satisfied by the core loading pattern. Based on the core loading pattern, nuclear, and thermal and hydraulic characteristics are analyzed. Also included are fuel performance analysis and development of a core protection and monitoring system that is adequate for the soluble boron free core of an integral reactor. Core shielding design analysis is accomplished, too. Moreover, full scope interface data are produced for reactor safety and performance analyses and other design activities. Nuclear, thermal and

  16. Low void effect (CFV) core concept flexibility: from self-breeder to burner core - 15091

    International Nuclear Information System (INIS)

    Buiron, L.; Dujcikova, L.

    2015-01-01

    In the frame of the French strategy on sustainable nuclear energy, several scenarios consider fuel cycle transition toward a plutonium multi-recycling strategy in sodium cooled fast reactor (SFR). Basically, most of these scenarios consider the deployment of a 60 GWe SFR fleet in 2 steps to renew the French PWR fleet. As scenarios do investigate long term deployment configurations, some of them require tools for nuclear phase-out studies. Instead of designing new reactors, the adopted strategy does focus on adaptation of existing ones into burner configurations. This is what was done in the frame of the EFR project at the end of the 90's using the CAPRA approach (French acronym for Enhance Plutonium Consumption in Fast Reactor). The EFR burner configuration was obtained by inserting neutronic penalties inside the core (absorber material and/or diluent subassembly). Starting from the preliminary industrial image of a SFR 3600 MWth core based on Low Sodium Void concept (CFV in French), a 'CAPRA-like' approach has been studied. As the CFV self-breeding is ensured by fertile blankets, a first modification consisted in the substitution of the corresponding depleted uranium by 'inert' or absorber material leading to a 'natural burner' core with only small impacts on flux distribution. The next step forward CAPRA configuration was the substitution of 1/3 of the fuel pins by 'dummy' pins (MgO pellets). The small spectrum shift due to MgO material insertion leads to an increase Doppler constant which exceeds the value of the reference case. As the core sodium void worth value is conserved, the CFV CAPRA core 'safety' potential is quite similar to the one of the reference core. Fuel thermo-mechanical requirements are met by both nominal core power and fuel time residence reduction. However, these reduction factors are lower than those obtained for EFR core. The management of the enhanced reactivity swing is discussed

  17. IVVS probe mechanical concept design

    Energy Technology Data Exchange (ETDEWEB)

    Rossi, Paolo, E-mail: paolo.rossi@enea.it; Neri, Carlo; De Collibus, Mario Ferri; Mugnaini, Giampiero; Pollastrone, Fabio; Crescenzi, Fabio

    2015-10-15

    Highlights: • ENEA designed, developed and tested a laser based In Vessel Viewing System (IVVS). • IVVS mechanical design has been revised from 2011 to 2013 to meet ITER requirements. • Main improvements are piezoceramic actuators and a step focus system. • Successful qualification activities validated the concept design for ITER environment. - Abstract: ENEA has been deeply involved in the design, development and testing of a laser based In Vessel Viewing System (IVVS) required for the inspection of ITER plasma-facing components. The IVVS probe shall be deployed into the vacuum vessel, providing high resolution images and metrology measurements to detect damages and possible erosion. ENEA already designed and manufactured an IVVS probe prototype based on a rad-hard concept and driven by commercial micro-step motors, which demonstrated satisfying viewing and metrology performances at room conditions. The probe sends a laser beam through a reflective rotating prism. By rotating the axes of the prism, the probe can scan all the environment points except those present in a shadow cone and the backscattered light signal is then processed to measure the intensity level (viewing) and the distance from the probe (metrology). During the last years, in order to meet all the ITER environmental conditions, such as high vacuum, gamma radiation lifetime dose up to 5 MGy, cumulative neutron fluence of about 2.3 × 10{sup 17} n/cm{sup 2}, temperature of 120 °C and magnetic field of 8 T, the probe mechanical design was significantly revised introducing a new actuating system based on piezo-ceramic actuators and improved with a new step focus system. The optical and mechanical schemes have been then modified and refined to meet also the geometrical constraints. The paper describes the mechanical concept design solutions adopted in order to fulfill IVVS probe functional performance requirements considering ITER working environment and geometrical constraints.

  18. Information science and its core concepts

    DEFF Research Database (Denmark)

    Hjørland, Birger

    2013-01-01

    One often encounters disagreements in information science (IS) (or library and information science, LIS), even disagreements about what might seem rather trivial questions. Such disagreements range from the designation of the field to questions such as whether IS is an academic discipline or not...... terminological hygiene” may account for some of the disagreements, but basically the problem is seen as a lack of sufficient strong centripetal tendencies keeping the field together....

  19. Analysis of digester design concepts

    Energy Technology Data Exchange (ETDEWEB)

    Ashare, E.; Wilson, E. H.

    1979-01-29

    Engineering economic analyses were performed on various digester design concepts to determine the relative performance for various biomass feedstocks. A comprehensive literature survey describing the state-of-the-art of the various digestion designs is included. The digester designs included in the analyses are CSTR, plug flow, batch, CSTR in series, multi-stage digestion and biomethanation. Other process options investigated included pretreatment processes such as shredding, degritting, and chemical pretreatment, and post-digestion processes, such as dewatering and gas purification. The biomass sources considered include feedlot manure, rice straw, and bagasse. The results of the analysis indicate that the most economical (on a unit gas cost basis) digester design concept is the plug flow reactor. This conclusion results from this system providing a high gas production rate combined with a low capital hole-in-the-ground digester design concept. The costs determined in this analysis do not include any credits or penalties for feedstock or by-products, but present the costs only for conversion of biomass to methane. The batch land-fill type digester design was shown to have a unit gas cost comparable to that for a conventional stirred tank digester, with the potential of reducing the cost if a land-fill site were available for a lower cost per unit volume. The use of chemical pretreatment resulted in a higher unit gas cost, primarily due to the cost of pretreatment chemical. A sensitivity analysis indicated that the use of chemical pretreatment could improve the economics provided a process could be developed which utilized either less pretreatment chemical or a less costly chemical. The use of other process options resulted in higher unit gas costs. These options should only be used when necessary for proper process performance, or to result in production of a valuable by-product.

  20. Web-based Core Design System Development

    International Nuclear Information System (INIS)

    Moon, So Young; Kim, Hyung Jin; Yang, Sung Tae; Hong, Sun Kwan

    2011-01-01

    The selection of a loading pattern is one of core design processes in the operation of a nuclear power plant. A potential new loading pattern is identified by selecting fuels that to not exceed the major limiting factors of the design and that satisfy the core design conditions for employing fuel data from the existing loading pattern of the current operating cycle. The selection of a loading pattern is also related to the cycle plan of an operating nuclear power plant and must meet safety and economic requirements. In selecting an appropriate loading pattern, all aspects, such as input creation, code runs and result processes are processed as text forms manually by a designer, all of which may be subject to human error, such as syntax or running errors. Time-consuming results analysis and decision-making processes are the most significant inefficiencies to avoid. A web-based nuclear plant core design system was developed here to remedy the shortcomings of an existing core design system. The proposed system adopts the general methodology of OPR1000 (Korea Standard Nuclear Power Plants) and Westinghouse-type plants. Additionally, it offers a GUI (Graphic User Interface)-based core design environment with a user-friendly interface for operators. It reduces human errors related to design model creation, computation, final reload core model selection, final output confirmation, and result data validation and verification. Most significantly, it reduces the core design time by more than 75% compared to its predecessor

  1. Core design methods for advanced LMFBRs

    International Nuclear Information System (INIS)

    Chandler, J.C.; Marr, D.R.; McCurry, D.C.; Cantley, D.A.

    1977-05-01

    The multidiscipline approach to advanced LMFBR core design requires an iterative design procedure to obtain a closely-coupled design. HEDL's philosophy requires that the designs should be coupled to the extent that the design limiting fuel pin, the design limiting duct and the core reactivity lifetime should all be equal and should equal the fuel residence time. The design procedure consists of an iterative loop involving three stages of the design sequence. Stage 1 consists of general mechanical design and reactor physics scoping calculations to arrive at an initial core layout. Stage 2 consists of detailed reactor physics calculations for the core configuration arrived at in Stage 1. Based upon the detailed reactor physics results, a decision is made either to alter the design (Stage 1) or go to Stage 3. Stage 3 consists of core orificing and detailed component mechanical design calculations. At this point, an assessment is made regarding design adequacy. If the design is inadequate the entire procedure is repeated until the design is acceptable

  2. Study on core design for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Okubo, Tsutomu

    2002-01-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  3. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  4. Conceptual core designs for a 1200 MWe sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

    2008-01-01

    The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

  5. Design of full MOX core in ABWR

    International Nuclear Information System (INIS)

    Kinoshita, Y.; Hirose, T.; Sasagawa, M.; Sakuma, T

    1999-01-01

    A Full MOX-ABWR, loaded with mixed-oxide (MOX) fuels of up to 100% of the core, is planned. Increased MOX fuel utilization will result in greater savings of uranium. Studies on the fuel rod thermal-mechanical design, the core design and the safety evaluation have been made, and the results are summarized in this paper. To sum it all up, the safety of the Full MOX-ABWR has been confirmed through design evaluations adequately considering the MOX fuel and core characteristics. (author)

  6. Core concepts for ''zero-sodium-void-worth core'' in metal fuelled fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.; Hill, R.N.; Fujita, E.K.; Wade, D.C.; Kumaoka, Y.; Suzuki, M.; Kawashima, M.; Nakagawa, H.

    1991-01-01

    Core design options to reduce the sodium void worth in metal fueled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a ''pancaked'' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. 16 refs., 2 figs., 3 tabs

  7. Core concepts for 'zero-sodium-void-worth core' in metal fuelled fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.; Hill, R.N.; Fujita, E.K.; Wade, D.C.; Kumaoka, Y.; Suzuki, M.; Kawashima, M.; Nakagawa, H.

    1991-01-01

    Core design options to reduce the sodium void worth in metal fuelled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a 'pancaked' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket-zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. (author)

  8. Core design and fuel management studies

    International Nuclear Information System (INIS)

    Min, Byung Joo; Chan, P.

    1997-06-01

    The design target for the CANDU 9 requires a 20% increase in electrical power output from an existing 480-channel CANDU core. Assuming a net electrical output of 861 MW(e) for a natural uranium fuelled Bruce-B/Darlington reactor in a warm water site, the net electrical output of the reference CANDU 9 reactor would be 1033 MW(e). This report documents the result of the physics studies for the design of the CANDU 9 480/SEU core. The results of the core design and fuel management studies of the CANDU 9 480/SEU reactor indicated that up to 1033 MW(e) output can be achieved in a 480-channel CANDU core by using SEU core can easily be maintained indefinitely using an automated refuelling program. Fuel performance evaluation based on the data of the 500 FPDs refuelling simulation concluded that SEU fuel failure is not expected. (author). 2 tabs., 38 figs., 5 refs

  9. Reactor core design aiding system

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro; Hamaguchi, Yukio; Nakao, Takashi; Kondo, Yasuhide

    1995-01-01

    A two-dimensional radial power distribution and an axial one-dimensional power distribution are determined based on a distribution of a three-dimensional infinite multiplication factor, to obtain estimated power distribution estimation values. The estimation values are synthesized to obtain estimated three-dimensional power distribution values. In addition, the distribution of a two-dimensional radial multiplication factor and the distribution of an one-dimensional axial multiplication factor are determined based on the three-dimensional power distribution, to obtain estimated values for the multiplication factor distribution. The estimated values are synthesized to form estimated values for the three-dimensional multiplication factor distribution. Further, estimated fuel loading pattern value is determined based on the three-dimensional power distribution or the two-dimensional radial power distribution. Since the processes for determining the estimated values comprise only additive and multiplying operations, processing time can be remarkably saved compared with calculation based on a detailed physical models. Since the estimation is performed on every fuel assemblies, a nervous circuit network not depending on the reactor core system can be constituted. (N.H.)

  10. Preliminary core design of IRIS-50

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Franceschini, Fausto

    2009-01-01

    IRIS-50 is a small, 50 MWe, advanced PWR with integral primary system. It evolved employing the same design principles as the well known medium size (335 MWe) IRIS. These principles include the 'safety-by-design' philosophy, simple and robust design, and deployment flexibility. The 50 MWe design addresses the needs of specific applications (e.g., power generation in small regional grids, water desalination and biodiesel production at remote locations, autonomous power source for special applications, etc.). Such applications may favor or even require longer refueling cycles, or may have some other specific requirements. Impact of these requirements on the core design and refueling strategy is discussed in the paper. Trade-off between the cycle length and other relevant parameters is addressed. A preliminary core design is presented, together with the core main reactor physics performance parameters. (author)

  11. Weaving latino cultural concepts into Preparedness Core Competency training.

    Science.gov (United States)

    Riley-Jacome, Mary; Parker, Blanca Angelica Gonzalez; Waltz, Edward C

    2014-01-01

    The New York • New Jersey Preparedness and Emergency Response Learning Center (NY•NJ PERLC) is one of 14 Centers funded by the Centers for Disease Control and Prevention designed to address the preparedness and response training and education needs of the public health workforce. One of the important niches, or focus areas for the Center, is training to improve the capacity of public health workers to respond with competence to the needs of vulnerable populations. During every phase of a disaster, racial and ethnic minorities, including Latinos, suffer worse outcomes than the general population. Communities with diverse cultural origins and limited English speakers often present more complex issues during public health emergencies. Training that incorporates cultural concepts into the Preparedness Core Competencies may improve the ability of public health workers to engage the Latino community in preparedness activities and ultimately improve outcomes during disasters. This article describes initiatives undertaken by the NY•NJ PERLC to improve the capacity of the public health workforce to respond competently to the needs of Latino populations. In 2012, the Center collaborated with national, state, and local partners to develop a nationwide broadcast founded on the Preparedness Core Competencies, Latinos During Emergencies: Cultural Considerations Impacting Disaster Preparedness. The widely viewed broadcast (497 sites in 47 states and 13 nations) highlighted the commonalities and differences within Latino culture that can impact emergency preparedness and response and outlined practical strategies to enhance participation. The success of the broadcast spurred a number of partner requests for training and technical assistance. Lessons learned from these experiences, including our "undercover" work at local Points of Dispensing, are incorporated into subsequent interactive trainings to improve the competency of public health workers. Participants recommended

  12. CORD, PWR Core Design and Fuel Management

    International Nuclear Information System (INIS)

    Trkov, Andrej

    1996-01-01

    1 - Description of program or function: CORD-2 is intended for core design applications of pressurised water reactors. The main objective was to assemble a core design system which could be used for simple calculations (such as frequently required for fuel management) as well as for accurate calculations (for example, core design after refuelling). 2 - Method of solution: The calculations are performed at the cell level with a lattice code in the supercell approximation to generate the single cell cross sections. Fuel assembly cross section homogenization is done in the diffusion approximation. Global core calculations can be done in the full three-dimensional cartesian geometry. Thermohydraulic feedbacks can be accounted for. The Effective Diffusion Homogenization method is used for generating the homogenized cross sections. 3 - Restrictions on the complexity of the problem: The complexity of the problem is selected by the user, depending on the capacity of his computer

  13. Design concept of KALIMER-600

    International Nuclear Information System (INIS)

    Hahn, Dohee; Kim, Yeong-Il; Kim, Seong-O; Lee, Jae-Han; Lee, Yong-Bum

    2005-01-01

    KALIMER-600 is a pool-type sodium-cooled reactor loaded with U-TRU-10%Zr metal fuels generating the net electricity output of 600 MWe. In order to enhance the proliferation resistance, no blanket assemblies are loaded in the core. To suppress the high power peaking factor, some of the fuel rods are replaced with B 4 C rods and dummy rods. The heat transport system is comprised of two independent loops of IHTS and SGS and the safety-grade residual heat removal system, PDRC, is a completely passive system. Main features of the mechanical structure design of KALIMER-600 are the seismically isolated reactor building, the reduced total pipe length of the IHTS, the simplified reactor support, and the compact reactor internal structures. From the safety analyses, the KALIMER-600 design is verified to be capable of accommodating all the analyzed ATWS events. This self-regulation capability of the KALIMER-600 is mainly due to the inherent reactivity feedback mechanisms and completely passive PDRC system. (author)

  14. Design of Concept Libraries for C++

    KAUST Repository

    Sutton, Andrew; Stroustrup, Bjarne

    2012-01-01

    algorithms and data structures and to gain insights into how best to support such concepts within C++. We start with the design of concepts rather than the design of supporting language features; the language design must be made to fit the concepts, rather

  15. Participation in a Year-Long CURE Embedded into Major Core Genetics and Cellular and Molecular Biology Laboratory Courses Results in Gains in Foundational Biological Concepts and Experimental Design Skills by Novice Undergraduate Researchers†

    Science.gov (United States)

    Peteroy-Kelly, Marcy A.; Marcello, Matthew R.; Crispo, Erika; Buraei, Zafir; Strahs, Daniel; Isaacson, Marisa; Jaworski, Leslie; Lopatto, David; Zuzga, David

    2017-01-01

    This two-year study describes the assessment of student learning gains arising from participation in a year-long curriculum consisting of a classroom undergraduate research experience (CURE) embedded into second-year, major core Genetics and Cellular and Molecular Biology (CMB) laboratory courses. For the first course in our CURE, students used micro-array or RNAseq analyses to identify genes important for environmental stress responses by Saccharomyces cerevisiae. The students were tasked with creating overexpressing mutants of their genes and designing their own original experiments to investigate the functions of those genes using the overexpression and null mutants in the second CURE course. In order to evaluate student learning gains, we employed three validated concept inventories in a pretest/posttest format and compared gains on the posttest versus the pretest with student laboratory final grades. Our results demonstrated that there was a significant correlation between students earning lower grades in the Genetics laboratory for both years of this study and gains on the Genetics Concept Assessment (GCA). We also demonstrated a correlation between students earning lower grades in the Genetics laboratory and gains on the Introductory Molecular and Cell Biology Assessment (IMCA) for year 1 of the study. Students furthermore demonstrated significant gains in identifying the variable properties of experimental subjects when assessed using the Rubric for Experimental (RED) design tool. Results from the administration of the CURE survey support these findings. Our results suggest that a year-long CURE enables lower performing students to experience greater gains in their foundational skills for success in the STEM disciplines. PMID:28904646

  16. Preliminary core design calculations for the ACPR Upgrade

    International Nuclear Information System (INIS)

    Pickard, P.S.

    1976-01-01

    The goal of the Annular Core Pulse Reactor (ACPR) Upgrade design studies is to define a core configuration that provides a significant increase in pulse fluence and fission energy deposition. The reactor modification should provide as flat an energy deposition profile for experiments as feasible. The fuels examined in this study were UO 2 -BeO (5-15 w/o UO 2 ), UC-ZrC-C (200-500 mg U/cc) and U-ZrH 1.5 . The basic core concept examined was a two region core, - a high heat capacity inner core region surrounded by an outer U-ZrH 1.5 region. Survey core calculations utilizing 1D transport calculations and cross sections libraries derived from the ORNL-AMPX code examined relative fuel loadings, fuel temperatures, reactivity requirements and pulse performance improvement. Reference designs for all candidate fuels were defined utilizing 2D transport and Monte Carlo calculations. The performance implications of alternative core designs were also examined for the UO 2 -BeO and UC-ZrC-C fuel candidates. (author)

  17. Design and development of small and medium integral reactor core

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Chang, M. H.; Lee, C. C.; Song, J. S.; Cho, B. O.; Kim, K. Y.; Kim, S. J.; Park, S. Y.; Lee, K. B.; Lee, C. H.; Chun, T. H.; Oh, D. S.; In, W. K.; Kim, H. K.; Lee, C. B.; Kang, H. S.; Song, K. N.

    1997-07-01

    Recently, the role of small and medium size integral reactors is remarkable in the heat applications rather than the electrical generations. Such a range of possible applications requires extensive used of inherent safety features and passive safety systems. It also requires ultra-longer cycle operations for better plant economy. Innovative and evolutionary designs such as boron-free operations and related reactor control methods that are necessary for simple reactor system design are demanded for the small and medium reactor (SMR) design, which are harder for engineers to implement in the current large size nuclear power plants. The goals of this study are to establish preliminary design criteria, to perform the preliminary conceptual design and to develop core specific technology for the core design and analysis for System-integrated Modular Advanced ReacTor (SMART) of 330 MWt power. Based on the design criteria of the commercial PWR's, preliminary design criteria will be set up. Preliminary core design concept is going to be developed for the ultra-longer cycle and boron-free operation and core analysis code system is constructed for SMART. (author). 100 refs., 40 tabs., 92 figs

  18. GFR fuel and core pre-conceptual design studies

    International Nuclear Information System (INIS)

    Chauvin, N.; Ravenet, A.; Lorenzo, D.; Pelletier, M.; Escleine, J.M.; Munoz, I.; Bonnerot, J.M.; Malo, J.Y.; Garnier, J.C.; Bertrand, F.; Bosq, J.C.

    2007-01-01

    The revision of the GFR core design - plate type - has been undertaken since previous core presented at Global'05. The self-breeding searched for has been achieved with an optimized design ('12/06 E'). The higher core pressure drop was a matter of concern. First of all, the core coolability in natural circulation for pressurized conditions has been studied and preliminary plant transient calculations have been performed. The design and safety criteria are met but no more margin remains. The project is also addressing the feasibility and the design of the fuel S/A. The hexagonal shape together with the principle of closed S/A (wrapper tube) is kept. Ceramic plate type fuel element combines a high enough core power density (minimization of the Pu inventory) and plutonium and minor actinides recycling capabilities. Innovative for many aspects, the fuel element is central to the GFR feasibility. It is supported already by a significant R and D effort also applicable to a pin concept that is considered as the other fuel element of interest. This combination of fuel/core feasibility and performance analysis, safety dispositions and performances analysis will compose the 'GFR preliminary feasibility' which is a project milestone at the end of the year 2007. (authors)

  19. Automated Design and Optimization of Pebble-bed Reactor Cores

    International Nuclear Information System (INIS)

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

    2010-01-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  20. Core design study on reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Hiroshi, Akie; Yoshihiro, Nakano; Toshihisa, Shirakawa; Tsutomu, Okubo; Takamichi, Iwamura

    2002-01-01

    The conceptual core design study of reduced-moderation water reactors (RMWRs) with tight-pitched MOX-fuelled lattice has been carried out at JAERI. Several different RMWR core concepts based on both BWR and PWR have been proposed. All the core concepts meet with the aim to achieve both a conversion ratio of 1.0 or larger and negative void reactivity coefficient. As one of these RMWR concepts, the ABWR compatible core is also proposed. Although the conversion ratio of this core is 1.0 and the void coefficient is negative, the discharge burn-up of the fuel was about 25 GWd/t. By adopting a triangular fuel pin lattice for the reduction of moderator volume fraction and modifying axial Pu enrichment distribution, it was aimed to extend the discharge burn-up of ABWR compatible type RMWR. By using a triangular fuel lattice of smaller moderator volume fraction, discharge burn-up of 40 GWd/t seems achievable, keeping the high conversion ratio and the negative void coefficient. (authors)

  1. Core concept of fast power reactor with zero sodium void reactivity

    International Nuclear Information System (INIS)

    Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

    1991-01-01

    The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

  2. Full MOX core design in ABWR

    International Nuclear Information System (INIS)

    Ihara, Toshiteru; Mochida, Takaaki; Izutsu, Sadayuki; Fujimaki, Shingo

    2003-01-01

    Electric Power Development Co., Ltd. (EPDC) has been investigating an ABWR plant for construction at Oma-machi in Aomori Prefecture. The reactor, termed FULL MOX-ABWR will have its reactor core eventually loaded entirely with mixed-oxide (MOX) fuel. Extended use of MOX fuel in the plant is expected to play important roles in the country's nuclear fuel recycling policy. MOX fuel bundles will initially be loaded only to less than one-third of the reactor, but will be increased to cover its entire core eventually. The number of MOX fuel bundles in the core thus varies anywhere from 0 to 264 for the initial cycle and, 0 to 872 for equilibrium cycles. The safety design of the FULL MOX-ABWR briefly stated next considers any probable MOX loading combinations out of such MOX bundle usage scheme, starting from full UO 2 to full MOX cores. (author)

  3. Conceptual Design of the RHIC Dump Core

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, A. J. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    1995-09-26

    Conceptually, the internal dump consists of a "core" whose purpose is to absorb the energy of the beam, and surrounding shielding whose purpose is to attenuate radiation. Design of the core for an internal dump has two problems which must be overcome. The first problem is preserving the integrity of the dump core. The bunches must be dispersed laterally an amount sufficient to keep the energy density from cracking the dump core material. Since the dump kickers in RHIC are only ~25m upstream of the entrance face of the dump, this is i a difficult problem. The second problem, not addressed in this note, is that dumping the beam should not quench downstream magnets. Preliminary calculations related to both of these problems have been presented in earlier notes.

  4. Multi-objective and multi-physics optimization methodology for SFR core: application to CFV concept

    International Nuclear Information System (INIS)

    Fabbris, Olivier

    2014-01-01

    Nuclear reactor core design is a highly multidisciplinary task where neutronics, thermal-hydraulics, fuel thermo-mechanics and fuel cycle are involved. The problem is moreover multi-objective (several performances) and highly dimensional (several tens of design parameters).As the reference deterministic calculation codes for core characterization require important computing resources, the classical design method is not well suited to investigate and optimize new innovative core concepts. To cope with these difficulties, a new methodology has been developed in this thesis. Our work is based on the development and validation of simplified neutronics and thermal-hydraulics calculation schemes allowing the full characterization of Sodium-cooled Fast Reactor core regarding both neutronics performances and behavior during thermal hydraulic dimensioning transients.The developed methodology uses surrogate models (or meta-models) able to replace the neutronics and thermal-hydraulics calculation chain. Advanced mathematical methods for the design of experiment, building and validation of meta-models allows substituting this calculation chain by regression models with high prediction capabilities.The methodology is applied on a very large design space to a challenging core called CFV (French acronym for low void effect core) with a large gain on the sodium void effect. Global sensitivity analysis leads to identify the significant design parameters on the core design and its behavior during unprotected transient which can lead to severe accidents. Multi-objective optimizations lead to alternative core configurations with significantly improved performances. Validation results demonstrate the relevance of the methodology at the pre-design stage of a Sodium-cooled Fast Reactor core. (author) [fr

  5. Concept Car Design and Ability Training

    Science.gov (United States)

    Lv, Jiefeng; Lu, Hairong

    The concept design as a symbol of creative design thinking, reflecting on the future design of exploratory and prospective, as a vehicle to explore the notion of future car design, design inspiration and creativity is not only a bold display, more through demonstrate the concept, reflects the company's technological strength and technological progress, and thus enhance their brand image. Present Chinese automobile design also has a very big disparity with world level, through cultivating students' concept design ability, to establish native design features and self-reliant brand image is practical and effective ways, also be necessary and pressing.

  6. Advance of core design method for ATR

    International Nuclear Information System (INIS)

    Maeda, Seiichirou; Ihara, Toshiteru; Iijima, Takashi; Seino, Hideaki; Kobayashi, Tetsurou; Takeuchi, Michio; Sugawara, Satoru; Matsumoto, Mitsuo.

    1995-01-01

    Core characteristics of ATR demonstration plant has been revised such as increasing the fuel burnup and the channel power, which is achieved by changing the number of fuel rod per fuel assembly from 28 to 36. The research and development concerning the core design method for ATR have been continued. The calculational errors of core analysis code have been evaluated using the operational data of FUGEN and the full scale simulated test results in DCA (Deuterium Critical Assembly) and HTL (Heat Transfer Loop) at O-arai engineering center. It is confirmed that the calculational error of power distribution is smaller than the design value of ATR demonstration plant. Critical heat flux correlation curve for 36 fuel rod cluster has been developed and the probability evaluation method based on its curve, which is more rational to evaluate the fuel dryout, has been adopted. (author)

  7. Design concept of Hydro cascade control system

    International Nuclear Information System (INIS)

    Fustik, Vangel; Kiteva, Nevenka

    2006-01-01

    In this paper a design concept of the comple hydro cascade scheme is presented with the design parameters of the main technical features. The cascade control system architecture is designed considering up-to-date communication and information technology. The control algorithm is based on Pond Level Control and Economic Load Allocation concepts.

  8. Fast breeder physics and nuclear core design

    International Nuclear Information System (INIS)

    Marth, W.; Schroeder, R.

    1983-07-01

    This report gathers the papers that have been presented on January 18/19, 1983 at a seminar ''Fast breeder physics and nuclear core design'' held at KfK. These papers cover the results obtained within about the last five years in the r+d program and give some indication, what still has to be done. To begin with, the ''tools'' of the core designer, i.e. nuclear data and neutronics codes are covered in a comprehensive way, the seminar emphasized the applications, however. First of all the accuracies obtained for the most important parameters are presented for the design of homogeneous and heterogeneous cores of about 1000 MWe, they are based on the results of critical experiments. This is followed by a survey on activities related to the KNK II reactor, i.e. calculations concerning a modification of the core as well as critical experiments done with respect to re-loads. Finally, work concerning reactivity worths of accident configurations is presented: the generation of reactivity worths for the input of safety-related calculations of a SNR 2 design, and critical experiments to investigate the requirements for the codes to be used for these calculations. These papers are accompanied by two contributions from the industrial partners. The first one deals with the requirements to nuclear design methods as seen by the reactor designer and then shows what has been achieved. The latter one presents state, trends, and methods of the SNR 2 design. The concluding remarks compare the state of the art reached within DeBeNe with international achievements. (orig.) [de

  9. Comparative study for minor actinide transmutation in various fast reactor core concepts

    International Nuclear Information System (INIS)

    Ohki, S.

    2001-01-01

    A comparative evaluation of minor actinide (MA) transmutation property was performed for various fast reactor core concepts. The differences of MA transmutation property were classified by the variations of fuel type (oxide, nitride, metal), coolant type (sodium, lead, carbon dioxide) and design philosophy. Both nitride and metal fuels bring about 10% larger MA transmutation amount compared with oxide fuel. The MA transmutation amount is almost unchanged by the difference between sodium and lead coolants, while carbon dioxide causes a reduction by about 10% compared with those. The changes of MA transmutation property by fuel and coolant types are comparatively small. The effects caused by the difference of core design are rather significant. (author)

  10. A CONCEPT OF SOLAR TRACKER SYSTEM DESIGN

    OpenAIRE

    Meita Rumbayan *, Muhamad Dwisnanto Putro

    2017-01-01

    Improvement of solar panel efficiency is an ongoing research work recently. Maximizing the output power by integrating with the solar tracker system becomes a interest point of the research. This paper presents the concept in designing a solar tracker system applied to solar panel. The development of solar panel tracker system design that consist of system display prototype design, hardware design, and algorithm design. This concept is useful as the control system for solar tracker to improve...

  11. Core design aspects of SNR 2

    International Nuclear Information System (INIS)

    Wehmann, U.K.

    1987-01-01

    The paper describes in its first part the main characteristics of the core of the SNR 2 fast breeder reactor which is being planned within the European collaboration on fast breeder reactors. In the second part some core design aspects are discussed. The fuel element management with an inwards shuffling after each cycle is illustrated which offers advantages with respect to linear rating, steel damage and average discharge burnup. For this management, the full three-dimensional power and burnup history has been calculated and some typical results are presented. The shutdown requirements and the capabilities of the two shutdown systems of SNR 2 are discussed. The necessity for a reliable surveillance of the power distribution is demonstrated by the pronounced power tilts in case of the unintentional withdrawal of an absorber rod. Finally, a short review of the main nuclear design methods and their validation with help of the evaluation of experiments in zero power facilities and power reactors is given

  12. Graphite core design in UK reactors

    International Nuclear Information System (INIS)

    Davies, M.W.

    1996-01-01

    The cores in the first power producing Magnox reactors in the UK were designed with only a limited amount of information available regarding the anisotropic dimensional change behaviour of Pile Grade graphite. As more information was gained it was necessary to make modifications to the design, some minor, some major. As the cores being built became larger, and with the switch to the Advanced Gas-cooled Reactor (AGR) with its much higher power density, additional problems had to be overcome such as increased dimensional change and radiolytic oxidation by the carbon dioxide coolant. For the AGRs a more isotropic graphite was required, with a lower initial open pore volume and higher strength. Gilsocarbon graphite was developed and was selected for all the AGRs built in the UK. Methane bearing coolants are used to limit radiolytic oxidation. (author). 5 figs

  13. Thermal hydraulics and mechanics core design programs

    International Nuclear Information System (INIS)

    Heinecke, J.

    1992-10-01

    The report documents the work performed within the Research and Development Task T hermal hydraulics and mechanics core design programs , funded by the German government. It contains the development of new codes, the extension of existing codes, the qualification and verification of codes and the development of a code library. The overall goal of this work was to adapt the system of thermal hydraulics and mechanics codes to the permanently growing requirements of the status of science and technology

  14. On the Design Concept in Engineering Ethics

    Science.gov (United States)

    Ohishi, Toshihiro

    The purpose of this study is to clarify the meaning of the trendy concept in engineering ethics education that ethical problems should be comprehended from the viewpoint of design. First, I present two objections against the concept and the content of it. Second, I examine the concept and show that the essence of it is pragmatic methods. That is, we should understand ethical problems and design problems pragmatically. Finally, I point out that the objections are not true of this pragmatic understanding.

  15. Thermal hydraulic design of PFBR core

    International Nuclear Information System (INIS)

    Roychowdhury, D.G.; Vinayagam, P.P.; Ravichandar, S.C.

    2000-01-01

    The thermal-hydraulic design of core is important in respecting temperature limits while achieving higher outlet temperature. This paper deals with the analytical process developed and implemented for analysing steady state thermal-hydraulics of PFBR core. A computer code FLONE has been developed for optimisation of flow allocation through the subassemblies (SA). By calibrating β n (ratio between the maximum channel temperature rise and SA average temperature rise) values with SUPERENERGY code and using these values in FLONE code, prediction of average and maximum coolant temperature distribution is found to be reasonably accurate. Hence, FLONE code is very powerful design tool for core design. A computer code SAPD has been developed to calculate the pressure drop of fuel and blanket SA. Selection of spacer wire pitch depends on the pressure drop, flow-induced vibration and the mixing characteristics. A parametric study was made for optimisation of spacer wire pitch for the fuel SA. Experimental programme with 19 pin-bundle has been undertaken to find the flow-induced vibration characteristics of fuel SA. Also, experimental programme has been undertaken on a full-scale model to find the pressure drop characteristics in unorificed SA, orifices and the lifting force on the SA. (author)

  16. ETHICAL FASHION CONCEPT AND DESIGNERS

    OpenAIRE

    Pinar GOKLUBERK OZLU; Kenan SAATCIOGLU

    2015-01-01

    Some problems like rapidly developing industrialization, irregular population growth, environmental pollution and to feel the impact of global warming as seriously, has been giving significant damage to the earth. People has realized that, after polluting to clean is harder than polluting of the measures to be taken before. And again people showed the sensitivity to the environment through different reactions and sanctions, took measures and created the new concepts about the enviroment. "Eth...

  17. Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2017-09-15

    In the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. Purpose of this article is twofold, on one hand to introduce a geometry modification in order to avoid the disadvantages of the original geometry and discuss new measurement results. On the other hand to present an analysis in order to suggest a method of proper numerical modelling of the problem based on the comparison of calculation results and measurement data for the new, modified geometry. The investigated concept has a homogeneous cylindrical core without any internal structures. Previous measurements on the scaled and segmented plexiglas model of the concept core and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong undesired flow separation could develop, that could negatively affect the characteristics of the core from neutronics point of view as well. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Particle image velocimetry (PIV) measurement results of the modified experimental model will be presented and compared to numerical simulation results with the purpose of CFD model validation.

  18. Application of the robust design concept for fuel loading pattern

    International Nuclear Information System (INIS)

    Endo, Tomohiro; Ohori, Kazuma; Yamamoto, Akio

    2011-01-01

    Application of the robust design concept for fuel loading pattern design is proposed as a new approach to improve the prediction accuracy of core characteristics. The robust design is a design concept that establishes a resistant (robust) system for perturbations or noises, by properly setting design variables. In order to apply the concept of robust design to fuel loading pattern design, we focus on a theoretical approach based on the higher order perturbation method. This approach indicates that the eigenvalue separation is one of the effective indices to measure the robustness of a designed fuel loading pattern. In order to verify the effectiveness of the eigenvalue separation as an index of robustness, numerical analysis is carried out for typical 3-loop PWR cores, and we evaluated the correlation between the eigenvalue separation and the variation of relative assembly power due to the perturbation of the cross section. The numerical results show that the variation of relative power decreases as the eigenvalue separation increases; thus, it is confirmed that the eigenvalue separation is an effective index of robustness. Based on the eigenvalue separation of a fuel loading pattern, we discuss design guidelines of a fuel loading pattern to improve the robustness. For example, if each fuel assembly has independent uncertainty on its cross section, the robustness of the core can be enhanced by increasing the relative power at the center of the core. The proposed guidelines will be useful to design a loading pattern that has robustness for uncertainties due to cross section, calculation method, and so on. (author)

  19. Core designs of modern VVER projects

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Kushmanov, S.; Vjalitsyn, V.; Vasilchenko, R.

    2015-01-01

    The presented operational experience of TVS - 2M (pilot-commercial operation started in 2006 at Balakovo NPP -1) enables to use it as reference for new projects because of similarity in designs and operational conditions. In the paper main parameters of fuel cycles, stability to impact of damaging factors, pilot operation of MG, new alloys, ADF and NTMC, upgrade of FA - 2M for the further power uprating, profiling of Gd-fuel rods for 18-month Fuel Cycle (FC) and perfection of absorber element design are the discussed issues. At the end author concluded that: 1) Core designs of new projects AES-2006 and VVER-TOI are based on extensive successful operational experience of the close prototype of TVS - 2M. 2) All improvements both of technical and economic parameters of fuel are subjected to representative examination by pilot operation at the power units with VVER-1000 being close prototypes of new designs

  20. SMART core preliminary nuclear design-II

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong Chan; Ji, Seong Kyun; Chang, Moon Hee

    1997-06-01

    Three loading patterns for 330 MWth SMART core are constructed for 25, 33 and 29 CRDMs, and one loading pattern for larger 69-FA core with 45 CRDMs is also constructed for comparison purpose. In this study, the core consists of 57 reduced height Korean Optimized Fuel Assemblies (KOFAs) developed by KAERI. The enrichment of fuel is 4.95 w/o. As a main burnable poison, 35% B-10 enriched B{sub 4}C-Al{sub 2}O{sub 3} shim is used. To control stuck rod worth, some gadolinia bearing fuel rods are used. The U-235 enrichment of the gadolinia bearing fuel rods is 1.8 w/o as used in KOFA. All patterns return cycle length of about 3 years. Three loading patterns except 25-CRDM pattern satisfy cold shutdown condition of keff {<=} 0.99 without soluble boron. These three patterns also satisfy the refueling condition of keff {<=} 0.95. In addition to the construction of loading pattern, an editing module of MASTER PPI files for rod power history generation is developed and rod power histories are generated for 29-CRDM loading pattern. Preliminary Fq design limit is suggested as 3.71 based on KOFA design experience. (author). 9 tabs., 45 figs., 16 refs.

  1. Computer-Aided Test Flow in Core-Based Design

    OpenAIRE

    Zivkovic, V.; Tangelder, R.J.W.T.; Kerkhoff, Hans G.

    2000-01-01

    This paper copes with the test-pattern generation and fault coverage determination in the core based design. The basic core-test strategy that one has to apply in the core-based design is stated in this work. A Computer-Aided Test (CAT) flow is proposed resulting in accurate fault coverage of embedded cores. The CAT now is applied to a few cores within the Philips Core Test Pilot IC project

  2. Design of Concept Libraries for C++

    KAUST Repository

    Sutton, Andrew

    2012-01-01

    We present a set of concepts (requirements on template arguments) for a large subset of the ISO C++ standard library. The goal of our work is twofold: to identify a minimal and useful set of concepts required to constrain the library\\'s generic algorithms and data structures and to gain insights into how best to support such concepts within C++. We start with the design of concepts rather than the design of supporting language features; the language design must be made to fit the concepts, rather than the other way around. A direct result of the experiment is the realization that to simply and elegantly support generic programming we need two kinds of abstractions: constraints are predicates on static properties of a type, and concepts are abstract specifications of an algorithm\\'s syntactic and semantic requirements. Constraints are necessary building blocks of concepts. Semantic properties are represented as axioms. We summarize our approach: concepts = constraints + axioms. This insight is leveraged to develop a library containing only 14 concepts that encompassing the functional, iterator, and algorithm components of the C++ Standard Library (the STL). The concepts are implemented as constraint classes and evaluated using Clang\\'s and GCC\\'s Standard Library test suites. © 2012 Springer-Verlag.

  3. Core design characteristics of the hyper system

    International Nuclear Information System (INIS)

    Yonghee, Kim; Won-Seok, Park; Hill, R.N.

    2003-01-01

    In Korea, an accelerator-driven system (ADS) called HYPER (Hybrid Power Extraction Reactor) is being studied for the transmutation of the radioactive wastes. HYPER is a 1000 MWth lead-bismuth eutectic (LBE)-cooled ADS. In this paper, the neutronic design characteristics of HYPER are described and its transmutation performances are assessed for an equilibrium cycle. The core is loaded with a ductless fuel assembly containing transuranics (TRU) dispersion fuel pins. In HYPER, a relatively high core height, 160 cm, is adopted to maximize the multiplication efficiency of the external source. In the ductless fuel assembly, 13 non-fuel rods are used as tie rods to maintain the mechanical integrity of assembly. As the reflector material, pure lead is used to improve the neutron economy and to minimise the generation of radioactive materials. In HYPER, to minimise the burn-up reactivity swing, a B 4 C burnable absorber is employed. For efficient depletion of the B-10 absorber, the burnable absorber is loaded only in the axially-central part (92 cm long) of the 13 tie rods of each assembly. In the current design, the amount of the B 4 C absorber was determined such that the burn-up reactivity swing is about 3.0% Δk. The long-lived fission products (LLFPs) 99 Tc and 129 I are also transmuted in the HYPER core such that their supporting ratios are equal to that of the TRUs. A heterogeneous LLFP transmutation in the reflector zone has been analysed in this work. A unique feature of the HYPER system is that it has an auxiliary core shutdown system, independent of the accelerator shutdown system. It has been shown that a cylindrical B 4 C absorber between the target and fuel blanket can drastically reduce the fission power even without shutting off the accelerator power. (author)

  4. Preliminary study on flexible core design of super FBR with multi-axial fuel shuffling

    International Nuclear Information System (INIS)

    Sukarman; Yamaji, Akifumi; Someya, Takayuki; Noda, Shogo

    2017-01-01

    Preliminary study has been conducted on developing a new flexible core design concept for the Supercritical water-cooled Fast Breeder Reactor (Super FBR) with multi-axial fuel shuffling. The proposed new concept focuses on the characteristic large axial coolant density change in supercritical water cooled reactors (SCWRs) when the coolant inlet temperature is below the pseudocritical point and large coolant enthalpy rise is taken in the core for achieving high thermal efficiency. The aim of the concept is to attain both the high breeding performance and good thermal-hydraulic performance at the same time. That is, short Compound System Doubling Time (CSDT) for high breeding, large coolant enthalpy rise for high thermal efficiency, and large core power. The proposed core concept consists of horizontal layers of mixed oxide (MOX) fuels and depleted uranium (DU) blanket layers at different elevation levels. Furthermore, the upper core and the lower core are separated and independent fuel shuffling schemes in these two core regions are considered. The number of fuel batches and fuel shuffling scheme of the upper core were changed to investigate influence of multi-axial fuel shuffling on the core characteristics. The core characteristics are evaluated with-three-dimensional diffusion calculations, which are fully-coupled with thermal-hydraulics calculations based on single channel analysis model. The results indicate that the proposed multi-axial fuel shuffling scheme does have a large influence on CSDT. Further investigations are necessary to develop the core concept. (author)

  5. Physicians’ Professionally Responsible Power: A Core Concept of Clinical Ethics

    Science.gov (United States)

    McCullough, Laurence B.

    2016-01-01

    The gathering of power unto themselves by physicians, a process supported by evidence-based practice, clinical guidelines, licensure, organizational culture, and other social factors, makes the ethics of power—the legitimation of physicians’ power—a core concept of clinical ethics. In the absence of legitimation, the physician’s power over patients becomes problematic, even predatory. As has occurred in previous issues of the Journal, the papers in the 2016 clinical ethics issue bear on the professionally responsible deployment of power by physicians. This introduction explores themes of physicians’ power in papers from an international group of authors who address autonomy and trust, the virtues of perinatal hospice, conjoined twins in ethics and law, addiction and autonomy in clinical research on addicting substances, euthanasia of patients with dementia in Belgium, and a pragmatic approach to clinical futility. PMID:26671961

  6. The materials challenge for LFR core design

    International Nuclear Information System (INIS)

    Grasso, Giacomo; Agostini, Pietro

    2013-01-01

    LFR share the main issues of all Fast Reactors, while presenting specific issues due to the use of lead as coolant. A number of constraints impairs the design of a LFR core, possibly resulting in a viability domain not exploitable for producing electricity in an efficient (hence economic) way. In particular, the most restrictive issues to be faced pend on the cladding. The selection of proper cladding materials provides the solution for the issues impairing the resistance of the cladding against stresses and irradiation effects. On the other hand, the protection of the cladding requires surface protections like oxide scales (passivation) or adherent layers (coating). Oxide scales seem not sufficient for a stable and effective protection of the base material. The application of adherent layers seems the only promising solution for protecting the cladding against corrosion. For the short term (i.e.: ALFRED), advanced 15/15Ti with coating is the reference solution for the cladding, allowing a core design complying with all the design constraints and goals. The candidate coatings are already being tested under irradiation to proceed towards qualification. In parallel, new base materials and/or coatings are presently under investigation. For the long term (i.e.: ELFR), the availability of such advanced materials/coatings might allow the extension of the viability domain towards higher and broader ranges (temperature, dpa, etc.), extending the fields of applications of LFRs and resulting in higher performances

  7. Preliminary design concepts for the advanced neutron source reactor systems

    International Nuclear Information System (INIS)

    Peretz, F.J.

    1988-01-01

    This paper describes the initial design work to develop the reactor systems hardware concepts for the advanced neutron source (ANS) reactor. This project has not yet entered the conceptual design phase; thus, design efforts are quite preliminary. This paper presents the collective work of members of the Oak Ridge National Laboratory, Martin Marietta Energy Systems, Inc., Engineering Division, and other participating organizations. The primary purpose of this effort is to show that the ANS reactor concept is realistic from a hardware standpoint and to show that project objectives can be met. It also serves to generate physical models for use in neutronic and thermal-hydraulic core design efforts and defines the constraints and objectives for the design. Finally, this effort will develop the criteria for use in the conceptual design of the reactor

  8. A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

    International Nuclear Information System (INIS)

    Yang, W.S.; Kim, T.K.; Grandy, C.

    2007-01-01

    This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% Δk. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% Δk. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

  9. Design evaluation of emergency core cooling systems using Axiomatic Design

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Gyunyoung [Massachusetts Institute of Technology, Department of Mechanical Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)]. E-mail: gheo@mit.edu; Lee, Song Kyu [Korea Advanced Institute of Science and Technology, Department of Nuclear and Quantum Engineering, 373-1 Guseong-dong, Yuseong-gu, Daejeon (Korea, Republic of)

    2007-01-15

    In designing nuclear power plants (NPPs), the evaluation of safety is one of the important issues. As a measure for evaluating safety, this paper proposes a methodology to examine the design process of emergency core cooling systems (ECCSs) in NPPs using Axiomatic Design (AD). This is particularly important for identifying vulnerabilities and creating solutions. Korean Advanced Power Reactor 1400 MWe (APR1400) adopted the ECCS, which was improved to meet the stronger safety regulations than that of the current Optimized Power Reactor 1000 MWe (OPR1000). To improve the performance and safety of the ECCS, the various design strategies such as independency or redundancy were implemented, and their effectiveness was confirmed by calculating core damage frequency. We suggest an alternative viewpoint of evaluating the deployment of design strategies in terms of AD methodology. AD suggests two design principles and the visualization tools for organizing design process. The important benefit of AD is that it is capable of providing suitable priorities for deploying design strategies. The reverse engineering driven by AD has been able to show that the design process of the ECCS of APR1400 was improved in comparison to that of OPR1000 from the viewpoint of the coordination of design strategies.

  10. ISABELLE control system: design concepts

    International Nuclear Information System (INIS)

    Humphrey, J.W.

    1979-01-01

    ISABELLE is a Department of Energy funded proton accelerator/storage ring being built at Brookhaven National Laboratory (Upton, Long Island, New York). It is large (3.8 km circumference) and complicated (approx. 30,000 monitor and control variables). It is based on superconducting technology. Following the example of previous accelerators, ISABELLE will be operated from a single control center. The control system will be distributed and will incorporate a local computer network. An overview of the conceptual design of the ISABELLE control system will be presented

  11. Advanced PWR fuel design concepts

    International Nuclear Information System (INIS)

    Andersor, C.K.; Harris, R.P.; Crump, M.W.; Fuhrman, N.

    1987-01-01

    For nearly 15 years, Combustion Engineering has provided pressurized water reactor fuel with the features most suppliers are now introducing in their advanced fuel designs. Zircaloy grids, removable upper end fittings, large fission gas plenum, high burnup, integral burnable poisons and sophisticated analytical methods are all features of C-E standard fuel which have been well proven by reactor performance. C-E's next generation fuel for pressurized water reactors features 24-month operating cycles, optimal lattice burnable poisons, increased resistance to common industry fuel rod failure mechanisms, and hardware and methodology for operating margin improvements. Application of these various improvements offer continued improvement in fuel cycle economics, plant operation and maintenance. (author)

  12. Design Principles of Open Innovation Concept – Universal Design Viewpoint

    OpenAIRE

    Mustaquim, Moyen; Nyström, Tobias

    2013-01-01

    The concept of open innovation is becoming an increasingly popular topic of interest and seems to promise a lot in organizational development. However, to date there are no certain design principles that can be followed by organizations on how to use open innovation successfully. In this paper seven design principles of open innovation concept have been proposed. The derived principles are the outcome which is based on the principles of universal design. The open innovation design, based on t...

  13. Structural concepts and details for seismic design

    International Nuclear Information System (INIS)

    Johnson, M.W.; Smietana, E.A.; Murray, R.C.

    1991-01-01

    As a part of the DOE Natural Phenomena Hazards Program, a new manual has been developed, entitled UCRL-CR-106554, open-quotes Structural Concepts and Details for Seismic Design.close quotes This manual describes and illustrates good practice for seismic-resistant design

  14. Design Concepts. Teacher Edition. Marketing Education LAPs.

    Science.gov (United States)

    Hawley, Jana

    This learning activity packet is designed to help prepare students to acquire a competency: how to use design concepts in preparation for a career in the fashion industry. The unit consists of the competency, four objectives, suggested learning activities, transparency masters, and a pretest/posttest with answer keys. Activities include a…

  15. Core design and performance of small inherently safe LMRs

    International Nuclear Information System (INIS)

    Orechwa, Y.; Khalil, H.; Turski, R.B.; Fujita, E.K.

    1986-01-01

    Oxide and metal-fueled core designs at the 900 MWt level and constrained by a requirement for interchangeability are described. The physics parameters of the two cores studied here indicate that metal-fueled cores display attractive economic and safety features and are more flexible than are oxide cores in adapting to currently-changing deployment scenarios

  16. Physicians' Professionally Responsible Power: A Core Concept of Clinical Ethics.

    Science.gov (United States)

    McCullough, Laurence B

    2016-02-01

    The gathering of power unto themselves by physicians, a process supported by evidence-based practice, clinical guidelines, licensure, organizational culture, and other social factors, makes the ethics of power--the legitimation of physicians' power--a core concept of clinical ethics. In the absence of legitimation, the physician's power over patients becomes problematic, even predatory. As has occurred in previous issues of the Journal, the papers in the 2016 clinical ethics issue bear on the professionally responsible deployment of power by physicians. This introduction explores themes of physicians' power in papers from an international group of authors who address autonomy and trust, the virtues of perinatal hospice, conjoined twins in ethics and law, addiction and autonomy in clinical research on addicting substances, euthanasia of patients with dementia in Belgium, and a pragmatic approach to clinical futility. © The Author 2015. Published by Oxford University Press, on behalf of the Journal of Medicine and Philosophy Inc. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  17. In-core Instrument Subcritical Verification (INCISV) - Core Design Verification Method - 358

    International Nuclear Information System (INIS)

    Prible, M.C.; Heibel, M.D.; Conner, S.L.; Sebastiani, P.J.; Kistler, D.P.

    2010-01-01

    According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed in-core self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of confirming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at various rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the ex-core detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities. (authors)

  18. Extending Sociotechnical Design to Project Conception

    DEFF Research Database (Denmark)

    Kampf, Constance Elizabeth

    2011-01-01

    Project management processes offer specific sites for understanding the interplay of the social and the technical. This article focuses on the connection between knowledge and technology through knowledge communication processes, cultural & rhetorical contexts in projects, and the iterative process...... and the Aarhus School of Business, University of Aarhus, Denmark. The analysis demonstrates the potential of knowledge communication concepts for social technical design and highlights the cultural context of the designers as a key factor to consider in socio-technical design....

  19. Learn from the Core--Design from the Core

    Science.gov (United States)

    Ockerse, Thomas

    2012-01-01

    The current objective, object-oriented approach to design is questioned along with design education viewed as a job-oriented endeavor. Instead relational knowledge and experience in a holistic sense, both tacit and explicit, are valued along with an appreciation of the unique character of the student. A new paradigm for design education is…

  20. Overview of core designs and requirements/criteria for core restraint systems

    International Nuclear Information System (INIS)

    Sutherland, W.H.

    1984-09-01

    The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented

  1. Computer-Aided Test Flow in Core-Based Design

    NARCIS (Netherlands)

    Zivkovic, V.; Tangelder, R.J.W.T.; Kerkhoff, Hans G.

    2000-01-01

    This paper copes with the test-pattern generation and fault coverage determination in the core based design. The basic core-test strategy that one has to apply in the core-based design is stated in this work. A Computer-Aided Test (CAT) flow is proposed resulting in accurate fault coverage of

  2. Main engineering features driving design concept and engineering design constraints

    International Nuclear Information System (INIS)

    Saito, Ryusei; Kobayashi, Takeshi; Yamada, Masao

    1987-09-01

    Major engineering design philosophies are described, which are essential bases for an engineering design and may have significant impacts on a reactor design concept. Those design philosophies are classified into two groups, engineering design drivers and engineering design constraints. The design drivers are featured by the fact that a designer is free to choose and the choice may be guided by his opinion, such as coil system, a mechanical configuration, a tritium breeding scenario, etc.. The design constraints may follow a natural law or engineering limit, such as material strength, coil current density, and so on. (author)

  3. Main physics features driving design concept and physics design constraints

    International Nuclear Information System (INIS)

    Fujisawa, Noboru; Sugihara, Masayoshi; Yamamoto, Shin

    1987-07-01

    Major physics design philosophies are described, which are essential bases for a plasma design and may have significant impacts on a reactor design concept. Those design philosophies are classified into two groups, physics design drivers and physics design constraints. The design drivers are featured by the fact that a designer is free to choose and the choice may be guided by his opinion, such as ignition, a pulse length, an operation scenario, etc.. The design constraints may follow a physical law, such as plasma confinement, β-limit, density limit, and so on. (author)

  4. Neutronic design of the XT-ADS core

    International Nuclear Information System (INIS)

    Van den Eynde, G.

    2007-01-01

    The EUROTRANS project is an integrated project in the 6th European Framework Program in the context of Partitioning and Transmutation. The objective of this project is the step-wise approach to a European Transmutation Demonstration. This project aims to deliver an advanced design of a small-scale Accelerator Driven System (ADS), XT-ADS, as well as the conceptual design of a European Facility for Industrial Transmutation (EFIT). The partners of this project accepted to use the MYRRHA Draft-2 design file as a starting basis for the design of the short-term XT-ADS demonstration machine. Instead of starting from a blank page, this allowed optimising an existing design towards the needs of XT-ADS, and this within the accepted limits of the safety requirements. Many options have been revisited and the framework is now set up. The main two objectives of the XT-ADS machine are the following: to demonstrate the feasibility of the ADS concept and to perform as a multi-purpose irradiation facility. Special attention is paid to the possibility of testing fuel dedicated to transmutation of minor actinides and long-life fission products. During the demonstration phase, the core will be loaded with MOX fuel in a clean core configuration. Since the XT-ADS must be a representative prototype, it has to operate at a reasonable power, a minimum of 50 MWth was set in the objectives. After this phase, the core will house In-Pile-Sections of different types for irradiating material samples, new types of fuel pins. We aim to be able to provide irradiation conditions that are close to EFIT conditions so XT-ADS can be used as a test-bed for EFIT parts

  5. A comparative design study of PB-BI cooled reactor cores with forced and natural convection cooling

    International Nuclear Information System (INIS)

    Mizuno, Tomoyasu; Enuma, Yasuhiro; Tanji, Mikio

    2003-01-01

    A comparative core design study is performed on Pb-Bi cooled reactors with forced and natural convection (FC and NC) cooling. Major interests of the study are core performance and core safety features. The designed core concepts with nitride fuel achieve reasonable breeding capability. The results of unprotected event analyses such as UTOP and ULOF show that both of concepts have possible features to withstand unprotected events due to negative reactivity feedback by Doppler effect, control rod drive line expansion, etc. These results lead to a conclusion that both of concepts have possible capability as one of future promising core concepts. A FC cooling core concept has more advantage if fuel recycle viewpoint is emphasized. (author)

  6. Proposal for a advanced PWR core with adequate characteristics for passive safety concept

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto

    1999-01-01

    This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

  7. Design comparisons of TRU burner cores with similar sodium void worth

    International Nuclear Information System (INIS)

    Sang Ji, Kim; Young Il, Kim; Young Jin, Kim; Nam Zin, Cho

    2001-01-01

    This study summarizes the neutronic performance and fuel cycle behavior of five geometrically-different transuranic (TRU) burner cores with similar low sodium void reactivity. The conceptual cores encompass core geometries for annular, two-region homogeneous, dual pin type, pan-shaped and H-shaped cores. They have been designed with the same assembly specifications and managed to have similar end-of-cycle sodium void reactivities and beginning-of-cycle peak power densities through the changes in the core size and configuration. The requirement of low sodium void reactivity is shown to lead each design concept to characteristic neutronics performance and fuel cycle behavior. The H-/pan-shaped cores allow the core compaction as well as higher rate of TRU burning. (author)

  8. An explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Iyoku, Tatsuo; Ishihara, Masahiro; Toyota, Junji; Shiozawa, Shusaku

    1991-05-01

    The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessary to make clear the basic concept to evaluate the integrity of core components of HTTR. Therefore, considering the detailed design of core graphite structures such as fuel graphite blocks, etc. of HTTR, this report explicates the design code in detail about the concepts of stress and fatigue limits, integrity evaluation method of oxidized graphite components and thermal irradiation stress analysis method etc. (author)

  9. Level design concept, theory, and practice

    CERN Document Server

    Kremers, Rudolf

    2009-01-01

    Good or bad level design can make or break any game, so it is surprising how little reference material exists for level designers. Beginning level designers have a limited understanding of the tools and techniques they can use to achieve their goals, or even define them. This book is the first to use a conceptual and theoretical foundation to build such a set of practical tools and techniques. It is tied to no particular technology or genre, so it will be a useful reference for many years to come. Kremers covers many concepts universal to level design, such as interactivity, world building, im

  10. Crashworthy airframe design concepts: Fabrication and testing

    Science.gov (United States)

    Cronkhite, J. D.; Berry, V. L.

    1982-01-01

    Crashworthy floor concepts applicable to general aviation aircraft metal airframe structures were investigated. Initially several energy absorbing lower fuselage structure concepts were evaluated. Full scale floor sections representative of a twin engine, general aviation airplane lower fuselage structure were designed and fabricated. The floors featured an upper high strength platform with an energy absorbing, crushable structure underneath. Eighteen floors were fabricated that incorporated five different crushable subfloor concepts. The floors were then evaluated through static and dynamic testing. Computer programs NASTRAN and KRASH were used for the static and dynamic analysis of the floor section designs. Two twin engine airplane fuselages were modified to incorporate the most promising crashworthy floor sections for test evaluation.

  11. The "Core Concepts Plus" Paradigm for Creating an Electronic Textbook for Introductory Business and Economic Statistics

    Science.gov (United States)

    Haley, M. Ryan

    2013-01-01

    This paper describes a flexible paradigm for creating an electronic "Core Concepts Plus" textbook (CCP-text) for a course in Introductory Business and Economic Statistics (IBES). In general terms, "core concepts" constitute the intersection of IBES course material taught by all IBES professors at the author's university. The…

  12. Design concept of the HPLWR moderator flow path

    International Nuclear Information System (INIS)

    Koehly, Christina; Schulenberg, Thomas; Starflinger, Joerg

    2009-01-01

    The latest design concept of the High Performance Light Water Reactor (HPLWR) includes a thermal core in which supercritical water at 25 MPa inlet pressure is heated up from 280degC reactor inlet temperature to 500degC core exit temperature in three steps with intermediate coolant mixing to minimize peak cladding temperatures of the fuel rods. Prior to entering the first fuel assemblies, the coolant is used as moderator in water rods inside assemblies, in the gap volume between assembly boxes, as well as in the surrounding axial or radial reflectors. Even though assembly boxes and moderator rods are designed with a certain thermal insulation, heat is generated in the moderator water or transferred to it from the superheated steam inside assemblies, causing concern of natural convection phenomena with uncontrolled neutronic feedback on the core power distribution. Moreover, bypass flows of the moderator water need to be minimized at any thermal expansion of the reactor internal structures to avoid an unpredictable moderator mass flow. The design concept of the moderator flow path described in this paper is trying to overcome these problems. Downward flow of moderator water is limited to sub-cooled conditions, well below the pseudo-critical point of supercritical water. Dedicated orifices are foreseen to allow later correction of the mass flow split. The sealing concept accounts for larger thermal expansions of reactor components by using C-rings or bellows. A welded construction is preferred wherever possible to minimize leakage. The removable steam plenum is aligned at the extractable steam pipes to minimize thermal displacements at the sealing positions. The paper is showing several design details to illustrate the technical solutions. (author)

  13. Conceptual design study of LMFBR core with carbide fuel

    International Nuclear Information System (INIS)

    Tezuka, H.; Hojuyama, T.; Osada, H.; Ishii, T.; Hattori, S.; Nishimura, T.

    1987-01-01

    Carbide fuel is a hopeful candidate for demonstration FBR(DFBR) fuel from the plant cost reduction point of view. High thermal conductivity and high heavy metal content of carbide fuel lead to high linear heat rate and high breeding ratio. We have analyzed carbide fuel core characteristics and have clarified the concept of carbide fuel core. By survey calculation, we have obtained a correlation map between core parameters and core characteristics. From the map, we have selected a high efficiency core whose features are better than those of an oxide core, and have obtained reactivity coefficients. The core volume and the reactor fuel inventory are approximately 20% smaller, and the burn-up reactivity loss is 50% smaller compared with the oxide fuel core. These results will reduce the capital cost. The core reactivity coefficients are similar to the conventional oxide DFBR's. Therefore the carbide fuel core is regarded as safe as the oxide core. Except neutron fluence, the carbide fuel core has better nuclear features than the oxide core

  14. Neutronic Design of KALIMER-600 Core with Moderator Rods

    International Nuclear Information System (INIS)

    Ser Gi Hong; Sang Ji Kim; Hoon Song; Yeong Il Kim

    2004-01-01

    Recently, the liquid-metal reactor research team of the Korea Atomic Energy Research Institute (KAERI) designed a 600 MWe sodium-cooled, metallic fueled fast reactor meeting the goals of Generation-IV, such as economics and proliferation resistance. In this paper, the core design analysis and its performance are reported. The core is designed to have a conversion ratio slightly larger than unity with no blanket assemblies in order not to produce an excess amount of high grade plutonium and to have no need for external feeds of fissile materials. To mitigate the sodium void reactivity of the fuel-self-sufficient core with no blanket assemblies, several design changes from a reference core are tried; reduction of the active core height, annular type cores with central dummy assemblies, and the use of moderator (BeO or ZrH 2 ) rods. As a result of the analysis, it is found that of the considered designs the use of moderator rods for the softening of the core neutron spectrum is the best choice for reducing the sodium void worth with the smallest changes from the reference fuel and assembly designs. The core analysis shows that the sodium void reactivity is reduced by ∼2$ in comparison with the reference core and the core has a much more negative fuel temperature reactivity feedback in comparison with the reference core. (authors)

  15. Mechatronic Systems Design Methods, Models, Concepts

    CERN Document Server

    Janschek, Klaus

    2012-01-01

    In this textbook, fundamental methods for model-based design of mechatronic systems are presented in a systematic, comprehensive form. The method framework presented here comprises domain-neutral methods for modeling and performance analysis: multi-domain modeling (energy/port/signal-based), simulation (ODE/DAE/hybrid systems), robust control methods, stochastic dynamic analysis, and quantitative evaluation of designs using system budgets. The model framework is composed of analytical dynamic models for important physical and technical domains of realization of mechatronic functions, such as multibody dynamics, digital information processing and electromechanical transducers. Building on the modeling concept of a technology-independent generic mechatronic transducer, concrete formulations for electrostatic, piezoelectric, electromagnetic, and electrodynamic transducers are presented. More than 50 fully worked out design examples clearly illustrate these methods and concepts and enable independent study of th...

  16. Design review report for rotary mode core sample truck (RMCST) modifications for flammable gas tanks, preliminary design

    International Nuclear Information System (INIS)

    Corbett, J.E.

    1996-02-01

    This report documents the completion of a preliminary design review for the Rotary Mode Core Sample Truck (RMCST) modifications for flammable gas tanks. The RMCST modifications are intended to support core sampling operations in waste tanks requiring flammable gas controls. The objective of this review was to validate basic design assumptions and concepts to support a path forward leading to a final design. The conclusion reached by the review committee was that the design was acceptable and efforts should continue toward a final design review

  17. Core design of a high breeding fast reactor cooled by supercritical pressure light water

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Takayuki, E-mail: russell@ruri.waseda.jp; Yamaji, Akifumi

    2016-01-15

    Highlights: • Core design concept of supercritical light water cooled fast breeding reactor is developed. • Compound system doubling time (CSDT) is applied for considering an appropriate target of breeding performance. • Breeding performance is improved by reducing fuel rod diameter of the seed assembly. • Core pressure loss is reduced by enlarging the coolant channel area of the seed assembly. - Abstract: A high breeding fast reactor core concept, cooled by supercritical pressure light water has been developed with fully-coupled neutronics and thermal-hydraulics core calculations, which takes into account the influence of core pressure loss to the core neutronics characteristics. Design target of the breeding performance has been determined to be compound system doubling time (CSDT) of less than 50 years, by referring to the relationship of energy consumption and economic growth rate of advanced countries such as the G7 member countries. Based on the past design study of supercritical water cooled fast breeder reactor (Super FBR) with the concept of tightly packed fuel assembly (TPFA), further improvement of breeding performance and reduction of core pressure loss are investigated by considering different fuel rod diameters and coolant channel geometries. The sensitivities of CSDT and the core pressure loss with respect to major core design parameters have been clarified. The developed Super FBR design concept achieves fissile plutonium surviving ratio (FPSR) of 1.028, compound system doubling time (CSDT) of 38 years and pressure loss of 1.02 MPa with positive density reactivity (negative void reactivity). The short CSDT indicates high breeding performance, which may enable installation of the reactors at a rate comparable to energy growth rate of developed countries such as G7 member countries.

  18. Computer-Aided Test Flow in Core-Based Design

    NARCIS (Netherlands)

    Zivkovic, V.; Tangelder, R.J.W.T.; Kerkhoff, Hans G.

    2000-01-01

    This paper copes with the efficient test-pattern generation in a core-based design. A consistent Computer-Aided Test (CAT) flow is proposed based on the required core-test strategy. It generates a test-pattern set for the embedded cores with high fault coverage and low DfT area overhead. The CAT

  19. Characteristic features of the core design of high-temperature reactors

    International Nuclear Information System (INIS)

    Brandes, S.; Lohnert, G.

    1975-01-01

    Following a survey on the possible applications of the HTGR depending on the height of the gas exiting temperatures, the core design for both of the fuel element concepts 'sphere' and 'block' is dealt with. The particularities arising from the multiple refueling and the one-way fueling in the design for spherical fuel elements are discussed. (UA/LH) [de

  20. Educational Videogames: Concept, Design And Evaluation

    Science.gov (United States)

    Rohrlick, D.; Yang, A.; Kilb, D. L.; Ma, L.; Ruzic, R.; Peach, C. L.; Layman, C. C.

    2013-12-01

    Videogames have historically gained popularity thanks to their entertainment rather than their educational value. This may be due, in part, to the fact that many educational videogames present academic concepts in dry, quiz-like ways, without the visual experiences, interactivity, and excitement of non-educational games. The increasing availability of tools that allow designers to easily create rich experiences for players now makes it simpler than ever for educational game designers to generate the visual experiences, interactivity, and excitement that gamers have grown to expect. Based on data from our work, when designed effectively, educational games can engage players, teach concepts, and tear down the stereotype of the stuffy, boring educational game. Our team has been experimenting with different ways to present scientific and mathematical concepts to middle and high school students through engaging, interactive games. When designing a gameplay concept, we focus on what we want the player to learn and experience as well as how to maintain a learning environment that is fun and engaging. Techniques that we have found successful include the use of a series of fast-paced 'minigames,' and the use of a 'simulator' learning method that allows a player to learn by completing objectives similar to those completed by today's scientists. Formative evaluations of our games over the past year have revealed both design strengths and weaknesses. Based on findings from a systematic evaluation of game play with diverse groups, with data collected through in-person observations of game play, knowledge assessments, focus groups, interviews with players, and computer tracking of students' game play behavior, we have found that players are uniformly enthusiastic about the educational tools. At the same time, we find there is more work to be done to make our tools fully intuitive, and to effectively present complex mathematical and scientific concepts to learners from a wide

  1. Comparative design of structures concepts and methodologies

    CERN Document Server

    Lin, Shaopei

    2016-01-01

    This book presents comparative design as an approach to the conceptual design of structures. Primarily focusing on reasonable structural performance, sustainable development and architectural aesthetics, it features detailed studies of structural performance through the composition and de-composition of these elements for a variety of structures, such as high-rise buildings, long-span crossings and spatial structures. The latter part of the book addresses the theoretical basis and practical implementation of knowledge engineering in structural design, and a case-based fuzzy reasoning method is introduced to illustrate the concept and method of intelligent design. The book is intended for civil engineers, structural designers and architects, as well as senior undergraduate and graduate students in civil engineering and architecture. Shaopei Lin and Zhen Huang are both Professors at the Department of Civil Engineering, Shanghai Jiao Tong University, China.

  2. Gallium Electromagnetic (GEM) Thrustor Concept and Design

    Science.gov (United States)

    Polzin, Kurt A.; Markusic, Thomas E.

    2006-01-01

    We describe the design of a new type of two-stage pulsed electromagnetic accelerator, the gallium electromagnetic (GEM) thruster. A schematic illustration of the GEM thruster concept is given in Fig. 1. In this concept, liquid gallium propellant is pumped into the first stage through a porous metal electrode using an electromagneticpump[l]. At a designated time, a pulsed discharge (approx.10-50 J) is initiated in the first stage, ablating the liquid gallium from the porous electrode surface and ejecting a dense thermal gallium plasma into the second state. The presence of the gallium plasma in the second stage serves to trigger the high-energy (approx.500 I), send-stage puke which provides the primary electromagnetic (j x B) acceleration.

  3. Design Report for the core design of the first core Mark-Ia of the SNR-300

    International Nuclear Information System (INIS)

    Stanculescu, A.

    1984-05-01

    The report describes the first core Mark-Ia of the SNR-300 reactor and its different assembly types with their operational strategy. Methods, criteria and results of the neutron physical, thermal hydraulic and core mechanical design of the whole core and its assemblies are presented

  4. Design concept for vessels and heat exchangers

    International Nuclear Information System (INIS)

    Elfmann, W.; Ferrari, L.D.B.

    1981-01-01

    A design concept for vessels and heat exchangers against internal and external loads resulting from normal operation and accident is shown. A definition and explanation of the operating conditions and stress levels are given. A description of the type of analysis (stress, fatigue, deformation, stability, earthquake and vibration) is presented in detail, also including technical guidelines which are used for the vessels and heat exchangers and their individual structure parts. (Author) [pt

  5. Constructive and thermal design of a core fast discharge

    International Nuclear Information System (INIS)

    Schroer, H.

    1979-08-01

    The present study is concerned with the development and thermal design of a fast discharge system for balls for the PR 3000 MWsub(th) process heat reactor. The term 'fast discharge system for balls' denotes a very short-time discharge procedure of the entire core contents, i.e. the flowing out of the fuel elements due to gravity into a receiver tank underneath the prestressed-concrete vessel. From a safety-engineering point of view, the fast discharge system for balls constitutes an additional possibility of active decay heat removal, besides the multiply redundant and diversitary reactor protection system, serving to further reduce the remaining residual risk. A fast discharge system for balls, however, is to be used only in the event of all the other possibilities of active decay heat removal having failed and when the maximum permissible temperatures for particularly exposed primary circuit components have been reached. However, the application range of such a system is restricted exclusively to high-temperature reactors with spherical fuel elements; the procedure cannot be applied to other reactor systems because of the rigidly fixed position of the fuel elements inside the core and for reasons of fuel element geometry. Besides the purpose of application, the influence of in-core temperature development on the possible actuation of the fast discharge system is being described in particular detail. This is followed by a description of the structural and thermal design of three specific major components, i.e. the piping system, shut-off devices and fuel element receiver tank, which will have to be installed additionally for the implementation of a fast discharge system for balls as compared to previous plant concepts. (orig.) [de

  6. Feasibility study of ultra-long life fast reactor core concept - 028

    International Nuclear Information System (INIS)

    Kim, T.K.; Taiwo, T.A.

    2010-01-01

    An ultra-long life core concept is proposed targeting capital and operational cost reductions and ultra-high discharge burnup in a fast reactor system. The core concept is achieved by de-rating the power density and adopting annular core geometry to maintain criticality for more than 40 years without refueling. The ultra-long life core has a specific power of ∼10 MW/t and an average driver fuel discharge burnup of ∼300 GWd/t. It is assumed such ultra-high burnup fuel can be developed within an advanced fuel cycle program. Several benefits are expected from the ultra-long life core concept such as capital and operational cost reductions, low proliferation risk, and effectively holding LWR spent fuel without disposal until technologies for a closed nuclear fuel cycle are developed and deployed. As future work, safety analysis, development of the advanced core cooling methods, and comparative cost analysis are expected. (authors)

  7. Preliminary concept of a zero power nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  8. Preliminary concept of a zero power nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

    2011-01-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  9. Preliminary ALARA design concept for SMART

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyo Youn; Kim, Seung Nam; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

    1999-03-01

    SMART(System-integrated Modular Advanced ReacTor) is a space saving integral type nuclear rector with the thermal power of 330 MW. This report provides general design guide and authority in NSSS designs for SMART needed to maintain the occupational doses and doses to members of public ALARA to meet the regulatory requirements. Paragraph 20.1 of 10 CFR 20, ''Standards for Protection Against Radiation'', states that licensee should make every reasonable effort to maintain exposures to radiation as far below the limits specified in Part 20 as is reasonably achievable. The ALARA (as low as is reasonably achievable) principle is incorporated into Korean radiation protection law as paragraph one Article 97 of the Atomic Energy Act. (Jan. 1995). This ALARA Design Concept for SMART provides 1) description of the organization and responsibilities needed for upper level management support and authority in order for the implementation of ALARA, 2) guidance and procedures for design, review, and evaluation needed for SMART ALARA program implementation, 3) general design guidelines for SMART NSSS and BOP designers to implement ALARA principles in design stage, and 4) training and instruction requirement of SMART NSSS and BOP designers for the familiarization of ALARA principles to be implemented in NSSS designs. (Author). 4 refs., 1 tabs.

  10. Preliminary ALARA design concept for SMART

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyo Youn; Kim, Seung Nam; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

    1999-03-01

    SMART(System-integrated Modular Advanced ReacTor) is a space saving integral type nuclear rector with the thermal power of 330 MW. This report provides general design guide and authority in NSSS designs for SMART needed to maintain the occupational doses and doses to members of public ALARA to meet the regulatory requirements. Paragraph 20.1 of 10 CFR 20, ''Standards for Protection Against Radiation'', states that licensee should make every reasonable effort to maintain exposures to radiation as far below the limits specified in Part 20 as is reasonably achievable. The ALARA (as low as is reasonably achievable) principle is incorporated into Korean radiation protection law as paragraph one Article 97 of the Atomic Energy Act. (Jan. 1995). This ALARA Design Concept for SMART provides 1) description of the organization and responsibilities needed for upper level management support and authority in order for the implementation of ALARA, 2) guidance and procedures for design, review, and evaluation needed for SMART ALARA program implementation, 3) general design guidelines for SMART NSSS and BOP designers to implement ALARA principles in design stage, and 4) training and instruction requirement of SMART NSSS and BOP designers for the familiarization of ALARA principles to be implemented in NSSS designs. (Author). 4 refs., 1 tabs.

  11. Preliminary ALARA design concept for SMART

    International Nuclear Information System (INIS)

    Kim, Kyo Youn; Kim, Seung Nam; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

    1999-03-01

    SMART(System-integrated Modular Advanced ReacTor) is a space saving integral type nuclear rector with the thermal power of 330 MW. This report provides general design guide and authority in NSSS designs for SMART needed to maintain the occupational doses and doses to members of public ALARA to meet the regulatory requirements. Paragraph 20.1 of 10 CFR 20, ''Standards for Protection Against Radiation'', states that licensee should make every reasonable effort to maintain exposures to radiation as far below the limits specified in Part 20 as is reasonably achievable. The ALARA (as low as is reasonably achievable) principle is incorporated into Korean radiation protection law as paragraph one Article 97 of the Atomic Energy Act. (Jan. 1995). This ALARA Design Concept for SMART provides 1) description of the organization and responsibilities needed for upper level management support and authority in order for the implementation of ALARA, 2) guidance and procedures for design, review, and evaluation needed for SMART ALARA program implementation, 3) general design guidelines for SMART NSSS and BOP designers to implement ALARA principles in design stage, and 4) training and instruction requirement of SMART NSSS and BOP designers for the familiarization of ALARA principles to be implemented in NSSS designs. (Author). 4 refs., 1 tabs

  12. Extending Sociotechnical design to project conception

    DEFF Research Database (Denmark)

    Kampf, Constance

    2009-01-01

    between knowledge and technology through knowledge communication processes, cultural and rhetorical contexts. This connection is examined from a process point of view through the development of project goals and objectives to situate technology. The data comes from a Project Management course in which...... the students were asked to design and plan projects to situate a mobile phone game in the social context around a museum in Helsinki or their online course management system.   The paper traces the evolution of students' project goals and objectives with respect to knowledge communication theory, demonstrating...... the potential of knowledge communication concepts for socio-technical design processes, as well as the implications of socio-technical design processes in extending our understanding of knowledge communication. Keywords: Knowledge Communication, Knowledge Management, Socio-Technical Design, Project Management....

  13. Basic criticality relations for gas core design

    International Nuclear Information System (INIS)

    Tanner, J.E.

    1992-01-01

    Minimum critical fissile concentrations are calculated for U-233, U-235, Pu-239, and Am-242m mixed homogeneously with hydrogen at temperatures to 15,000K. Minimum critical masses of the same mixtures in a 1000 liter sphere are also calculated. It is shown that propellent efficiencies of a gas core fizzler engine using Am-242m as fuel would exceed those in a solid core engine as small as 1000L operating at 100 atmospheres pressure. The same would be true for Pu-239 and possibly U-233 at pressures of 1000 atm. or at larger volumes

  14. Criteria design of the CAREM 25 reactor's core: neutronic aspects

    International Nuclear Information System (INIS)

    Lecot, C.A.

    1990-01-01

    The criteria that guided the design, from the neutronic point of view, of the CAREM reactor's core were presented. The minimum set of objectives and general criteria which permitted the design of the particular systems constituting the CAREM 25 reactor's core is detailed and stated. (Author) [es

  15. Development of concept and neutronic calculation method for large LMFBR core

    International Nuclear Information System (INIS)

    Shirakata, K.; Ishikawa, M.; Ikegami, T.; Sanda, T.; Kaneto, K.; Kawashima, M.; Kaise, Y.; Shirakawa, M.; Hibi, K.

    1991-01-01

    Presented in this paper is the state of the art of reactor physics R and Ds for the development of concept and neutronic calculation method for large Liquid Metal Fast Breeder Reactor (LMFBR) core. Physics characteristics of concepts for mixed oxide (MOX) fueled large FBR core were investigated by a series of benchmark critical experiments. Next, an adequacy and accuracy of the current neutronic calculation method was assessed by the experiments analyses, and then neutronic prediction accuracies by the method were evaluated for physics characteristics of the large core. Concerns on core development were discussed in terms of neutronics. (author)

  16. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  17. Overview of core designs and requirements/criteria for core restraint systems

    International Nuclear Information System (INIS)

    Sutherland, W.H.

    1984-01-01

    The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System is presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented. (author)

  18. Development of conceptual nuclear design of 10MWt research reactor core

    International Nuclear Information System (INIS)

    Kim, M. H.; Lim, J. Y.; Win, Naing; Park, J. M.

    2008-03-01

    KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

  19. GNPS 18-months fuel cycles core thermal hydraulic design

    International Nuclear Information System (INIS)

    Liu Changwen; Zhou Zhou

    2002-01-01

    GNPS begins to implement the 18-month fuel cycles from the initial annual reload at cycle 9, thus the initial core thermal hydraulic design is not valid any more. The new critical heat flux (CHF) correlation, FC, which is developed by Framatome, is used in the design, and the generalized statistical methodology (GSM) instead of the initial deterministic methodology is used to determine the DNBR design limit. As the AFA 2G and AFA 3G are mixed loaded in the transition cycle, it will result that the minimum DNBR in the mixed core is less than that of AFA 3G homogenous core, the envelop mixed core DNBR penalty is given. Consequently the core physical limit for mixed core and equilibrium cycles, and the new over temperature ΔT overpower ΔT are determined

  20. Improving work control systems: The core team concept

    International Nuclear Information System (INIS)

    Jorgensen, M.D.; Simpson, W.W.

    1996-01-01

    The improved work control system at the Idaho Chemical Processing Plant minimizes review and approval time, maximizes field work time, and maintains full compliance with applicable requirements. The core team method gives ownership and accountability to knowledgeable individuals, and the teams use sophisticated scheduling techniques to improve information sharing and cost control and to establish accurate roll-up master schedules

  1. Core design studies on various forms of coolants and fuel materials. 2. Studies on liquid heavy metal and gas cooled cores, small cores and evaluation of 4-type cores

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Sakashita, Yoshiyuki; Naganuma, Masayuki; Takaki, Naoyuki; Mizuno, Tomoyasu; Ikegami, Tetsuo

    2001-01-01

    Alternative concepts to sodium cooled fast reactors, such as heavy metal liquid cooled reactors and gas cooled fast reactors were studied in Phase-1 of the feasibility studies, aiming at simplification of the system, high thermal efficiency and enhancing safety. Fuel and core specifications and nuclear characteristics were surveyed to meet the targets for commercialization of fast reactor cycle. Nuclear characteristics of small fast reactor cores were also surveyed from the perspective of the possibility of multi-purpose use and dispersed power stations. The key points of the design study for each concept in Phase-2 were summarized from the aspect of the screening of the candidates for FR commercialization. (author)

  2. Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Marie, N.; Prulhière, G.; Lecerf, J. [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France)

    2016-02-15

    Highlights: • Low void worth CFV and SFRv2 cores are compared for ASTRID pre-conceptual design. • Severe accident behaviour is assessed with a simplified calculation approach and tools. • Mitigation to limit reactivity inserted by core compaction is easier for CFV than for SFRv2 core. • When facing arbitrary reactivity ramps, CFV core would lead to lower energy release than SFRv2 core. • Time scale for core degradation is one order of magnitude larger for CFV than for SFRv2. - Abstract: The present paper is dedicated to the studies carried out during the first stage of the pre-conceptual design of the French demonstrator of fourth generation SFR reactors (ASTRID) in order to compare the behaviour of two envisaged core concepts under severe accident transients. Among the two studied core concepts, whose powers are 1500 MWth, the first one is a classical homogeneous core (called SFRv2) with large pin diameter whose the sodium overall voiding reactivity effect is 5 $. The second concept is an axially heterogeneous core (called CFV) whose global void reactivity effect is negative (−1.2 $ at the end of cycle at the equilibrium). The comparison of the cores relies on two typical accident families: a reactivity insertion (unprotected transient overpower, UTOP) and an overall loss of core cooling (unprotected loss of flow, ULOF). In the first part of the comparison, the primary phase of an UTOP is studied in order to assess typical features of the transient behaviour: power and reactivity evolutions, material heating and melting/vaporization and mechanical energy release due to fuel vapor expansion. The second part of the comparison deals with the calculation of the reactivity potential for degraded states (molten pools) representative of the secondary phase of a mild UTOP and of a strong UTOP (strong or mild qualifies the reactivity ramp inserted). According to the reactivity potential, the amount of fuel to extract from the core and the amount of absorber

  3. Retrievable storage concept designs. Final report

    International Nuclear Information System (INIS)

    Nickell, R.E.

    1979-01-01

    Three tasks related to the reference design of retrievable storage canisters for radioactive waste have been completed. The three tasks consist of the reference design itself, the definition of failure modes most appropriate for structural integrity determinations for the reference canister, and the development of a failure methodology for the structural integrity of the containers. The reference design is a sealed storage canister concept based upon the waste isolation pilot plant (WIPP) design, with slight modifications. The modifications consist of an alternate lifting yoke arrangement for the top head and a revised bottom head design for absorption of impact energy. Welded closures provide the seal at each end. Overpacking is considered as a possibility, but is not included in the preliminary reference design. The four failure modes that are deemed the most appropriate for the design of the reference canister are: (i) a loss of functional capability; (ii) ductile rupture of the canister; (iii) buckling of the structural members; and (iv) stress corrosion cracking. Failure scenarios are provided for each of the relevant failure modes. In addition, a failure methodology based upon the distribution of demand and the distribution of capacity for the structural members, with respect to each failure mode, is proffered

  4. Action Relations. Basic Design Concepts for Behaviour Modelling and Refinement.

    NARCIS (Netherlands)

    Quartel, Dick

    This thesis presents basic design concepts, design methods and a basic design language for distributed system behaviours. This language is based on two basic concepts: the action concept and the causality relation concept. Our methods focus on behaviour refinement, which consists of replacing an

  5. Current fusion power plant design concepts

    International Nuclear Information System (INIS)

    Gore, B.F.; Murphy, E.S.

    1976-09-01

    Nine current U.S. designs for fusion power plants are described in this document. Summary tabulations include a tenth concept, for which the design document was unavailable during preparation of the descriptions. The information contained in the descriptions was used to define an envelope of fusion power plant characteristics which formed the basis for definition of reference first commercial fusion power plant design. A brief prose summary of primary plant features introduces each of the descriptions contained in the body of this document. In addition, summary tables are presented. These tables summarize in side-by-side fashion, plant parameters, processes, combinations of materials used, requirements for construction materials, requirements for replacement materials during operation, and production of wastes

  6. How Cultural Knowledge Shapes Core Design Thinking

    DEFF Research Database (Denmark)

    Clemmensen, Torkil; Ranjan, Apara; Bødker, Mads

    2018-01-01

    The growing trend of co-creation and co-design in cross-cultural design teams presents challenges for the design thinking process. We integrate two frameworks, one on reasoning patterns in design thinking, the other on the dynamic constructivist theory of culture, to propose a situation specific...... framework for the empirical analysis of design thinking in cross-cultural teams. We illustrate the framework with a qualitative analysis of 16 episodes of design related conversations, which are part of a design case study. The results show that cultural knowledge, either as shared by the cross......-cultural team or group specific knowledge of some team members, shape the reasoning patterns in the design thinking process across all the 16 episodes. Most of the design discussions were approached by the designers as problem situations that were formulated in a backward direction, where the value to create...

  7. Heavy ion driven LMF design concept

    International Nuclear Information System (INIS)

    Lee, E.P.

    1991-08-01

    The USA Department of Energy has conducted a multi-year study of the requirements, designs and costs for a Laboratory Microfusion Facility (LMF). The primary purpose of the LMF would be testing of weapons physics and effects simulation using the output from microexplosions of inertial fusion pellets. It does not need a high repetition rate, efficient driver system as required by an electrical generating plant. However there would be so many features in common that the design, construction and operation of an LMF would considerably advance the application of inertial confinement fusion to energy production. The DOE study has concentrated particularly on the LMF driver, with design and component development undertaken at several national laboratories. Principally, these are LLNL (Solid State Laser), LANL (Gas Laser), and SNLA (Light Ions). Heavy Ions, although considered a possible LMF driver did not receive attention until the final stages of this study since its program management was through the Office of Energy Research rather than Defense Programs. During preparation of a summary report for the study it was decided that some account of heavy ions was needed for a complete survey of the driver candidates. A conceptual heavy ion LMF driver design was created for the DOE report which is titled LMC Phase II Design Concepts. The heavy ion driver did not receive the level of scrutiny of the other concepts and, unlike the others, no costs analysis by an independent contractor was performed. Since much of heavy ion driver design lore was brought together in this exercise it is worthwhile to make it available as an independent report. This is reproduced here as it appears in the DOE report

  8. A safety design approach for sodium cooled fast reactor core toward commercialization in Japan

    International Nuclear Information System (INIS)

    Kubo, Shigenobu

    2012-01-01

    JAEA’s safety approach for SFR core design is based on defence‐in‐depth concept, which includes DBAs and DECs (prevention and mitigation): • The reactor core is designed to have inherent reactivity feedback characteristics with negative power coefficient. • Operation temperature range is set sufficiently below the coolant boiling temperature so as to avoid coolant boiling against anticipated operational occurrences and DBAs. • If the plant state deviates from operational states, the safe reactor shutdown is achieved by automatic insertion of control rods. 2 active reactor shutdown systems are provided. • Failure of active reactor shutdown is assumed in a design extension condition . Passive shutdown capability is provided by SASS under such condition. • As a design extension condition, core disruptive accident is assumed. In order to prevent severe mechanical energy release which might cause containment function failure, core sodium void worth is limited below 6 dollars and molten fuel discharge capability is utilized by FAIDUS. (author)

  9. New Design Concept for Universal CCD Controller

    Directory of Open Access Journals (Sweden)

    Wonyong Han

    1994-06-01

    Full Text Available Currently, the CCDs are widely used in astronomical observations either in direct imaging use or spectroscopic mode. However according to the recent technical advances, new large format CCDs are rapidly developed which have better performances with higher quantum efficiency and sensitivity. In many cases, some microprocessors have been adopted to deal with necessary digital logic for a CCD imaging system. This could often lack the flexibility of a system for a user to upgrade with new devices, especially of it is a commercial product. A new design concept has been explored which could provide the opportunity to deal with any format of devices from ant manufactures effectively for astronomical purposes. Recently available PLD (Programmable Logic Devices technology makes it possible to develop such digital circuit design, which can be integrated into a single component, instead of using microprocessors. The design concept could dramatically increase the efficiency and flexibility of a CCD imaging system, particularly when new or large format devices are available and to upgrade the performance of a system. Some variable system control parameters can be selected by a user with a wider range of choice. The software can support such functional requirements very conveniently. This approach can be applied not only to astronomical purpose, but also to some related fields, such as remote sensing and industrial applications.

  10. Project Design Concept - Primary Ventilation System

    International Nuclear Information System (INIS)

    MCGREW, D.L.

    2000-01-01

    Tank Farm Restoration and Safe Operation (TFRSO), Project W-3 14 was established to provide upgrades that would improve the reliability and extend the system life of portions of the waste transfer, electrical, ventilation, instrumentation and control systems for the Hanford Site Tank Farms. An assessment of the tank farm system was conducted and the results are documented in system assessment reports. Based on the deficiencies identified in the tank farm system assessment reports, and additional requirements analysis performed in support of the River Protection Project (RPP), an approved scope for the TFRSO effort was developed and documented in the Upgrade Scope Summary Report (USSR), WHC-SD-W314-RPT-003, Rev. 4. The USSR establishes the need for the upgrades and identifies the specific equipment to be addressed by this project. This Project Design Concept (PDC) is in support of the Phase 2 upgrades and provides an overall description of the operations concept for the W-314 Primary Ventilation Systems. Actual specifications, test requirements, and procedures are not included in this PDC. The PDC is a ''living'' document, which will be updated throughout the design development process to provide a progressively more detailed description of the W-314 Primary Ventilation Systems design. The Phase 2 upgrades to the Primary Ventilation Systems shall ensure that the applicable current requirements are met for: Regulatory Compliance; Safety; Mission Requirements; Reliability; and Operational Requirements

  11. AP1000 core design with 50% MOX loading

    International Nuclear Information System (INIS)

    Fetterman, Robert J.

    2009-01-01

    The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core design and a mixed MOX/UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance

  12. AP1000 core design with 50% MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)], E-mail: fetterrj@westinghouse.com

    2009-04-15

    The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core design and a mixed MOX/UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.

  13. Identifying and Overcoming Threshold Concepts and Conceptions: Introducing a Conception-Focused Curriculum to Course Design

    Science.gov (United States)

    Burch, Gerald F.; Burch, Jana J.; Bradley, Thomas P.; Heller, Nathan A.

    2015-01-01

    Educators have been challenged to identify threshold concepts and develop transformed students. This stands in stark contrast to many curriculum design and delivery models that currently view students as repositories of knowledge. In this article, we argue that educators can reach both goals, identify stumbling blocks and transforming students,…

  14. Core reset system design for linear induction accelerator

    International Nuclear Information System (INIS)

    Durga Praveen Kumar, D.; Mitra, S.; Sharma, Archana; Nagesh, K.V.; Chakravarthy, D.P.

    2006-01-01

    A repetitive pulsed power system based Linear Induction Accelerator (LIA-200) is being developed at BARC to get an electron beam of 200keV, 5kA, 50ns, 10-100 Hz. Amorphous core is the heart of these accelerators. It serves various functions in different subsystems viz. pulse power modulator, pulse transformer, magnetic switches and induction cavities. One of the factors that make the magnetic components compact is utilization of the total flux swing available in the core. In the present system, magnetic switches, pulse transformers, and induction cavity are designed to avail the full flux swing available in the core. For achieving this objective, flux density in the core has to be kept at the reverse saturation, before the main pulse is applied. The electrical circuit which makes it possible is called the core reset system. In this paper the details of core reset system designed for LIA-200 are described. (author)

  15. Building Integrated Design Practice under the Concept of Sustainable Development

    Science.gov (United States)

    Liu, Xuexin

    2018-03-01

    With the continuous development of social economy, people are more demanding for architecture. Some advanced design concepts are gradually applied to the design of buildings. Under the concept of sustainable development, building integration design has also been widely used to promote the rapid development of architectural design. Integrated design concepts and sustainable development concepts play an important role to meet people’s requirements. This article will explore the concept of sustainable development under the concept of integrated architectural design and practice analysis, propose appropriate measures.

  16. A review of the core catcher design in LMR

    International Nuclear Information System (INIS)

    Lee, Yong Bum; Hahn, Do Hee

    2001-08-01

    The overwhelming emphasis in reactor safety is on the prevention of core meltdown. Moreover, although there have been several accidents that have resulted in some fuel melting, to date there have been no accidents severe enough to cause the syndrome of core collapse, reactor vessel melt-through, containment penetration, and dispersal into the ground. Nevertheless, a number of proposals have been made for the design of core catcher systems to control or stop the motion of the molten core mass should such an accident take place. Core catchers may differ in both their location within the reactor system and in the mechanism that is used to cool and control the motion of the core debris. In this report the classification, configuration and main features of the core catcher are described. And also, The core catcher design technologies and processes are presented. Finally the core catcher provisions in constructed and planned LMRs (Liquid Metal Reactors) are summarized and the preliminary assessment on the core catcher installation in KALIMER is presented

  17. Thermal-hydraulic mixing in the split-core ANS reactor design

    International Nuclear Information System (INIS)

    Dorning, R.J.J.

    1988-01-01

    A design has been proposed for the advanced neutron source (ANS) reactor that incorporates a split core, one purpose of which is to create a mixing plenum between the upper and lower cores. It was hoped that in addition to introducing various desirable neutronics features, such as decreasing the fast neutron flux contamination of thermal and cold neutron beams located in the reactor midplane, this mixing plenum would make possible higher operating powers by lowering the maximum core temperature. This lower temperature was to be achieved as a result of the mixing, of the hot D 2 O coolant exiting the upper-core channels, and the cold D 2 O leaving the large upper core bypass. It was expected that this mixing would bring about a significantly reduced lower core maximum coolant inlet temperature. The authors have carried out large-scale computer calculations to determine the extent to which this mixing occurs in current split-core design geometry, which does not incorporate baffles, mixing devices, or other design features introduced to enhance mixing. The large-scale self-consistent calculations summarized here indicate that innovative design ideas to enhance mixing will be necessary if the split-core concept is to achieve the amount of thermal mixing needed to make possible significantly higher power operation and corresponding higher flux sources

  18. Long-term plutonium storage: Design concepts

    International Nuclear Information System (INIS)

    Wilkey, D.D.; Wood, W.T.; Guenther, C.D.

    1994-01-01

    An important part of the Department of Energy (DOE) Weapons Complex Reconfiguration (WCR) Program is the development of facilities for long-term storage of plutonium. The WCR design goals are to provide storage for metals, oxides, pits, and fuel-grade plutonium, including material being held as part of the Strategic Reserve and excess material. Major activities associated with plutonium storage are sorting the plutonium inventory, material handling and storage support, shipping and receiving, and surveillance of material in storage for both safety evaluations and safeguards and security. A variety of methods for plutonium storage have been used, both within the DOE weapons complex and by external organizations. This paper discusses the advantages and disadvantages of proposed storage concepts based upon functional criteria. The concepts discussed include floor wells, vertical and horizontal sleeves, warehouse storage on vertical racks, and modular storage units. Issues/factors considered in determining a preferred design include operational efficiency, maintenance and repair, environmental impact, radiation and criticality safety, safeguards and security, heat removal, waste minimization, international inspection requirements, and construction and operational costs

  19. AP1000 core design with 50% MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)

    2008-07-01

    The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core and a mixed MOX / UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

  20. AP1000 core design with 50% MOX loading

    International Nuclear Information System (INIS)

    Fetterman, Robert J.

    2008-01-01

    The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core and a mixed MOX / UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

  1. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yoon, K. H.; Lee, C. B.

    2014-01-01

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness

  2. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  3. BioCore Guide: A Tool for Interpreting the Core Concepts of Vision and Change for Biology Majors

    Science.gov (United States)

    Freeman, Scott; Wenderoth, Mary Pat; Crowe, Alison J.

    2014-01-01

    Vision and Change in Undergraduate Biology Education outlined five core concepts intended to guide undergraduate biology education: 1) evolution; 2) structure and function; 3) information flow, exchange, and storage; 4) pathways and transformations of energy and matter; and 5) systems. We have taken these general recommendations and created a Vision and Change BioCore Guide—a set of general principles and specific statements that expand upon the core concepts, creating a framework that biology departments can use to align with the goals of Vision and Change. We used a grassroots approach to generate the BioCore Guide, beginning with faculty ideas as the basis for an iterative process that incorporated feedback from more than 240 biologists and biology educators at a diverse range of academic institutions throughout the United States. The final validation step in this process demonstrated strong national consensus, with more than 90% of respondents agreeing with the importance and scientific accuracy of the statements. It is our hope that the BioCore Guide will serve as an agent of change for biology departments as we move toward transforming undergraduate biology education. PMID:26086653

  4. Safety design concept and analysis for the upgrading JRR-3

    International Nuclear Information System (INIS)

    Onishi, N.; Isshiki, M.; Takahashi, H.; Takayanagi, M.

    1990-01-01

    The Research Reactor No.3 (JRR-3) is under reconstruction for upgrading. This paper describes the safety design concepts of the architectural and engineering design, anticipated operational transients and accident conditions which are the postulated initiating events for the safety evaluation, and the safety criteria of the upgraded JRR-3. The safety criteria are defined taking into account those of Light Water Reactors and the characteristics of the research reactor. Using the example of the safety analysis, this paper describes analytical results of a reactivity insertion by removal of in-core irradiation samples, a pipeline break at the primary coolant loop and flow blockage to a coolant channel, which are the severest postulated initiating events of the JRR-3

  5. The Study on the Core Concepts of Contemporary Sociology of Education and Its Theoretical Construction

    Science.gov (United States)

    Qian, Min-hui

    2006-01-01

    Within the sphere of contemporary social sciences, the terms "modernity," "post-modernity" and "globalization" have penetrated, as the core concepts, into various fields of social sciences in a logical way. In constituting the concept of "modernity," sociology of education develops the educational theory, as sociological theory does, into a "grand…

  6. Public Conceptions of Algorithms and Representations in the Common Core State Standards for Mathematics

    Science.gov (United States)

    Nanna, Robert J.

    2016-01-01

    Algorithms and representations have been an important aspect of the work of mathematics, especially for understanding concepts and communicating ideas about concepts and mathematical relationships. They have played a key role in various mathematics standards documents, including the Common Core State Standards for Mathematics. However, there have…

  7. Quality By Design: Concept To Applications.

    Science.gov (United States)

    Swain, Suryakanta; Padhy, Rabinarayan; Jena, Bikash Ranjan; Babu, Sitty Manohar

    2018-03-08

    Quality by Design is associated to the modern, systematic, scientific and novel approach which is concerned with pre-distinct objectives that not only focus on product, process understanding but also leads to process control. It predominantly signifies the design and product improvement and the manufacturing process in order to fulfill the predefined manufactured goods or final products quality characteristics. It is quite essential to identify desire and required product performance report such as Target Product Profile, typical Quality Target Product Profile (QTPP) and Critical Quality attributes (CQA). This review highlighted about the concepts of QbD design space, for critical material attributes (CMAs) as well as the critical process parameters that can totally affect the CQAs within which the process shall be unaffected and consistently manufacture the required product. Risk assessment tools and design of experiments are its prime components. This paper outlines the basic knowledge of QbD, the key elements; steps as well as various tools for QbD implementation in pharmaceutics field are presented briefly. In addition to this, quite a lot of applications of QbD in numerous pharmaceutical related unit operations are discussed and summarized. This article provides a complete data as well as the road map for universal implementation and application of QbD for pharmaceutical products. Copyright© Bentham Science Publishers; For any queries, please email at epub@benthamscience.org.

  8. Design Concepts for Muon-Based Accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Ryne, R. D. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Berg, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Kirk, H. G. [Brookhaven National Lab. (BNL), Upton, NY (United States); Palmer, R. B. [Brookhaven National Lab. (BNL), Upton, NY (United States); Stratkis, D. [Brookhaven National Lab. (BNL), Upton, NY (United States); Alexahin, Y. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Bross, A. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Gollwitzer, K. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Mokhov, N. V. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Neuffer, D. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Palmer, M. A. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Yonehara, K. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Snopok, P. [IIT, Chicago, IL (United States); Bogacz, A. [Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Roberts, T. J. [Muons Inc., Batavia, IL (United States); Delahaye, J. -P. [SLAC National Accelerator Lab., Menlo Park, CA (United States)

    2015-05-01

    Muon-based accelerators have the potential to enable facilities at both the Intensity and the Energy Frontiers. Muon storage rings can serve as high precision neutrino sources, and a muon collider is an ideal technology for a TeV or multi-TeV collider. Progress in muon accelerator designs has advanced steadily in recent years. In regard to 6D muon cooling, detailed and realistic designs now exist that provide more than 5 order-of-magnitude emittance reduction. Furthermore, detector performance studies indicate that with suitable pixelation and timing resolution, backgrounds in the collider detectors can be significantly reduced, thus enabling high-quality physics results. Thanks to these and other advances in design & simulation of muon systems, technology development, and systems demonstrations, muon storage-ring-based neutrino sources and a muon collider appear more feasible than ever before. A muon collider is now arguably among the most compelling approaches to a multi-TeV lepton collider. This paper summarizes the current status of design concepts for muon-based accelerators for neutrino factories and a muon collider.

  9. The Triton: Design concepts and methods

    Science.gov (United States)

    Meholic, Greg; Singer, Michael; Vanryn, Percy; Brown, Rhonda; Tella, Gustavo; Harvey, Bob

    1992-01-01

    During the design of the C & P Aerospace Triton, a few problems were encountered that necessitated changes in the configuration. After the initial concept phase, the aspect ratio was increased from 7 to 7.6 to produce a greater lift to drag ratio (L/D = 13) which satisfied the horsepower requirements (118 hp using the Lycoming O-235 engine). The initial concept had a wing planform area of 134 sq. ft. Detailed wing sizing analysis enlarged the planform area to 150 sq. ft., without changing its layout or location. The most significant changes, however, were made just prior to inboard profile design. The fuselage external diameter was reduced from 54 to 50 inches to reduce drag to meet the desired cruise speed of 120 knots. Also, the nose was extended 6 inches to accommodate landing gear placement. Without the extension, the nosewheel received an unacceptable percentage (25 percent) of the landing weight. The final change in the configuration was made in accordance with the stability and control analysis. In order to reduce the static margin from 20 to 13 percent, the horizontal tail area was reduced from 32.02 to 25.0 sq. ft. The Triton meets all the specifications set forth in the design criteria. If time permitted another iteration of the calculations, two significant changes would be made. The vertical stabilizer area would be reduced to decrease the aircraft lateral stability slope since the current value was too high in relation to the directional stability slope. Also, the aileron size would be decreased to reduce the roll rate below the current 106 deg/second. Doing so would allow greater flap area (increasing CL(sub max)) and thus reduce the overall wing area. C & P would also recalculate the horsepower and drag values to further validate the 120 knot cruising speed.

  10. Reactor Core Design and Analysis for a Micronuclear Power Source

    Directory of Open Access Journals (Sweden)

    Hao Sun

    2018-03-01

    Full Text Available Underwater vehicle is designed to ensure the security of country sea boundary, providing harsh requirements for its power system design. Conventional power sources, such as battery and Stirling engine, are featured with low power and short lifetime. Micronuclear reactor power source featured with higher power density and longer lifetime would strongly meet the demands of unmanned underwater vehicle power system. In this paper, a 2.4 MWt lithium heat pipe cooled reactor core is designed for micronuclear power source, which can be applied for underwater vehicles. The core features with small volume, high power density, long lifetime, and low noise level. Uranium nitride fuel with 70% enrichment and lithium heat pipes are adopted in the core. The reactivity is controlled by six control drums with B4C neutron absorber. Monte Carlo code MCNP is used for calculating the power distribution, characteristics of reactivity feedback, and core criticality safety. A code MCORE coupling MCNP and ORIGEN is used to analyze the burnup characteristics of the designed core. The results show that the core life is 14 years, and the core parameters satisfy the safety requirements. This work provides reference to the design and application of the micronuclear power source.

  11. Scalable Multi-core Architectures Design Methodologies and Tools

    CERN Document Server

    Jantsch, Axel

    2012-01-01

    As Moore’s law continues to unfold, two important trends have recently emerged. First, the growth of chip capacity is translated into a corresponding increase of number of cores. Second, the parallalization of the computation and 3D integration technologies lead to distributed memory architectures. This book provides a current snapshot of industrial and academic research, conducted as part of the European FP7 MOSART project, addressing urgent challenges in many-core architectures and application mapping.  It addresses the architectural design of many core chips, memory and data management, power management, design and programming methodologies. It also describes how new techniques have been applied in various industrial case studies. Describes trends towards distributed memory architectures and distributed power management; Integrates Network on Chip with distributed, shared memory architectures; Demonstrates novel design methodologies and frameworks for multi-core design space exploration; Shows how midll...

  12. CopperCore, an Open Source IMS Learning Design Engine

    NARCIS (Netherlands)

    Vogten, Hubert

    2004-01-01

    The presentation gives an overview of the approach of the development programme of the OTEC department towards the development of Open Source. The CopperCore IMS Learning Design engine is described as an example of this approach.

  13. Improvement of SSR core design for ABWR-II

    International Nuclear Information System (INIS)

    Moriwaki, Masanao; Aoyama, Motoo; Okada, Hiroyuki; Kitamura, Hideya; Sakurada, Koichi; Tanabe, Akira

    2003-01-01

    In order to enhance the spectral shift effect in the ABWR-II reactor, a novel core design to bring out better performance of spectral shift rods (SSRs) is studied. The SSR is a new type of water rod, in which the water level develops naturally during operation and changes according to the coolant flow rate through the channel. By using the SSR, the average moderator density, which is directly related to core reactivity, can be controlled over a wide range by the core flow rate. In the new SSR core design, two types of SSR bundles, in which settings for the SSR water levels are different, are utilized and loaded according to flow distribution in the core. This two-region SSR core design allows wide variation in the average SSR water level, thus improving fuel economy. Enhancement of SSR function in the two-region SSR core increases the uranium saving factor by about 25%, from the 6% of the conventional uniform SSR core to about 8%. (author)

  14. Borehole disposal design concept in Madagascar

    International Nuclear Information System (INIS)

    Randriamarolahy, J.N.; Randriantseheno, H.F.; Andriambololona, Raoelina

    2008-01-01

    Full text: In Madagascar, sealed radioactive sources are used in several socio-economic sectors such as medicine, industry, research and agriculture. At the end of their useful lives, these radioactive sources become ionizing radiations waste and can be still dangerous because they can cause harmful effects to the public and the environment. 'Borehole disposal design concept' is needed for sitting up a safe site for storage of radioactive waste, in particular, sealed radioactive sources. Borehole disposal is an option for long-term management of small quantities of radioactive waste in compliance with the internationally accepted principles for radioactive waste management. Several technical aspects must be respected to carry out such a site like the geological, geomorphologic, hydrogeology, geochemical, meteorological and demographic conditions. Two sites are most acceptable in Madagascar such as Ankazobe and Fanjakana. A Borehole will be drilled and constructed using standard techniques developed for water abstraction, oil exploration. At the Borehole, the sealed radioactive sources are encapsulated. The capsule is inserted in a container. This type of storage is benefit for the developing countries because it is technologically simple and economic. The construction cost depends on the volume of waste to store and the Borehole depth. The borehole disposal concept provides a good level of safety to avoid human intrusion. The future protection of the generations against the propagation of the ionizing radiations is then assured. (author)

  15. Deep Space Habitat ECLSS Design Concept

    Science.gov (United States)

    Curley, Su; Stambaugh, Imelda; Swickrath, Michael; Anderson, Molly S.; Rotter, Henry

    2012-01-01

    Life support is vital to human spaceflight, and most current life support systems employ single-use hardware or regenerable technologies that throw away the waste products, relying on resupply to make up the consumables lost in the process. Because the long-term goal of the National Aeronautics and Space Administration is to expand human presence beyond low-earth orbit, life support systems must become self-sustaining for missions where resupply is not practical. From May through October 2011, the life support team at the Johnson Space Center was challenged to define requirements, develop a system concept, and create a preliminary life support system design for a non-planetary Deep Space Habitat that could sustain a crew of four in near earth orbit for a duration of 388 days. Some of the preferred technology choices to support this architecture were passed over because the mission definition has an unmanned portion lasting 825 days. The main portion of the architecture was derived from technologies currently integrated on the International Space Station as well as upcoming technologies with moderate Technology Readiness Levels. The final architecture concept contains only partially-closed air and water systems, as the breakeven point for some of the closure technologies was not achieved with the mission duration.

  16. Deep Space Habitat ECLS Design Concept

    Science.gov (United States)

    Curley, Su; Stambaugh, Imelda; Swickrath, Mike; Anderson, Molly; Rotter, Hank

    2011-01-01

    Life support is vital to human spaceflight, and most current life support systems employ single-use hardware or regenerable technologies that throw away the waste products, relying on resupply to make up the consumables lost in the process. Because the long-term goal of the National Aeronautics and Space Administration is to expand human presence beyond low-earth orbit, life support systems must become self-sustaining for missions where resupply is not practical. From May through October 2011, the life support team at the Johnson Space Center was challenged to define requirements, develop a system concept, and create a preliminary life support system design for a non-planetary Deep Space Habitat that could sustain a crew of four in near earth orbit for a duration of 388 days. Some of the preferred technology choices to support this architecture were passed over as the mission definition also has an unmanned portion lasting 825 days. The main portion of the architecture was derived from technologies currently integrated on the International Space Station as well as upcoming technologies with moderate Technology Readiness Levels. The final architecture concept contains only partially-closed air and water systems, as the breakeven point for some of the closure technologies was not achieved with the mission duration.

  17. Action Relations. Basic Design Concepts for Behaviour Modelling and Refinement.

    OpenAIRE

    Quartel, Dick

    1998-01-01

    This thesis presents basic design concepts, design methods and a basic design language for distributed system behaviours. This language is based on two basic concepts: the action concept and the causality relation concept. Our methods focus on behaviour refinement, which consists of replacing an abstract behaviour by a more concrete behaviour, such that the concrete behaviour conforms to the abstract behaviour. An important idea underlying this thesis is that an effective design methodology s...

  18. Review on JMTR safety design for LEU core conversion

    International Nuclear Information System (INIS)

    Komori, Yoshihiro; Yokokawa, Makoto; Saruta, Toru; Inada, Seiji; Sakurai, Fumio; Yamamoto, Katsumune; Oyamada, Rokuro; Saito, Minoru

    1993-12-01

    Safety of the JMTR was fully reviewed for the core conversion to low enriched uranium fuel. Fundamental policies for the JMTR safety design were reconsidered based on the examination guide for safety design of test and research reactors, and safety of the JMTR was confirmed. This report describes the safety design of the JMTR from the viewpoint of major functions for reactor safety. (author)

  19. Reference repository design concept for bedded salt

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, D.W.; Martin, R.W.

    1980-10-08

    A reference design concept is presented for the subsurface portions of a nuclear waste repository in bedded salt. General geologic, geotechnical, hydrologic and geochemical data as well as descriptions of the physical systems are provided for use on generic analyses of the pre- and post-sealing performance of repositories in this geologic medium. The geology of bedded salt deposits and the regional and repository horizon stratigraphy are discussed. Structural features of salt beds including discontinuities and dissolution features are presented and their effect on repository performance is discussed. Seismic hazards and the potential effects of earthquakes on underground repositories are presented. The effect on structural stability and worker safety during construction from hydrocarbon and inorganic gases is described. Geohydrologic considerations including regional hydrology, repository scale hydrology and several hydrological failure modes are presented in detail as well as the hydrological considerations that effect repository design. Operational phase performance is discussed with respect to operations, ventilation system, shaft conveyances, waste handling and retrieval systems and receival rates of nuclear waste. Performance analysis of the post sealing period of a nuclear repository is discussed, and parameters to be used in such an analysis are presented along with regulatory constraints. Some judgements are made regarding hydrologic failure scenarios. Finally, the design and licensing process, consistent with the current licensing procedure is described in a format that can be easily understood.

  20. SAFARI optical system architecture and design concept

    Science.gov (United States)

    Pastor, Carmen; Jellema, Willem; Zuluaga-Ramírez, Pablo; Arrazola, David; Fernández-Rodriguez, M.; Belenguer, Tomás.; González Fernández, Luis M.; Audley, Michael D.; Evers, Jaap; Eggens, Martin; Torres Redondo, Josefina; Najarro, Francisco; Roelfsema, Peter

    2016-07-01

    SpicA FAR infrared Instrument, SAFARI, is one of the instruments planned for the SPICA mission. The SPICA mission is the next great leap forward in space-based far-infrared astronomy and will study the evolution of galaxies, stars and planetary systems. SPICA will utilize a deeply cooled 2.5m-class telescope, provided by European industry, to realize zodiacal background limited performance, and high spatial resolution. The instrument SAFARI is a cryogenic grating-based point source spectrometer working in the wavelength domain 34 to 230 μm, providing spectral resolving power from 300 to at least 2000. The instrument shall provide low and high resolution spectroscopy in four spectral bands. Low Resolution mode is the native instrument mode, while the high Resolution mode is achieved by means of a Martin-Pupplet interferometer. The optical system is all-reflective and consists of three main modules; an input optics module, followed by the Band and Mode Distributing Optics and the grating Modules. The instrument utilizes Nyquist sampled filled linear arrays of very sensitive TES detectors. The work presented in this paper describes the optical design architecture and design concept compatible with the current instrument performance and volume design drivers.

  1. Reference repository design concept for bedded salt

    International Nuclear Information System (INIS)

    Carpenter, D.W.; Martin, R.W.

    1980-01-01

    A reference design concept is presented for the subsurface portions of a nuclear waste repository in bedded salt. General geologic, geotechnical, hydrologic and geochemical data as well as descriptions of the physical systems are provided for use on generic analyses of the pre- and post-sealing performance of repositories in this geologic medium. The geology of bedded salt deposits and the regional and repository horizon stratigraphy are discussed. Structural features of salt beds including discontinuities and dissolution features are presented and their effect on repository performance is discussed. Seismic hazards and the potential effects of earthquakes on underground repositories are presented. The effect on structural stability and worker safety during construction from hydrocarbon and inorganic gases is described. Geohydrologic considerations including regional hydrology, repository scale hydrology and several hydrological failure modes are presented in detail as well as the hydrological considerations that effect repository design. Operational phase performance is discussed with respect to operations, ventilation system, shaft conveyances, waste handling and retrieval systems and receival rates of nuclear waste. Performance analysis of the post sealing period of a nuclear repository is discussed, and parameters to be used in such an analysis are presented along with regulatory constraints. Some judgements are made regarding hydrologic failure scenarios. Finally, the design and licensing process, consistent with the current licensing procedure is described in a format that can be easily understood

  2. Design study of prestressed rotor spar concept

    Science.gov (United States)

    Gleich, D.

    1980-01-01

    Studies on the Bell Helicopter 540 Rotor System of the AH-1G helicopter were performed. The stiffness, mass and geometric configurations of the Bell blade were matched to give a dynamically similar prestressed composite blade. A multi-tube, prestressed composite spar blade configuration was designed for superior ballistic survivability at low life cycle cost. The composite spar prestresses, imparted during fabrication, are chosen to maintain compression in the high strength cryogenically stretchformed 304-L stainless steel liner and tension in the overwrapped HTS graphite fibers under operating loads. This prestressing results in greatly improved crack propagation and fatigue resistance as well as enhanced fiber stiffness properties. Advantages projected for the prestressed composite rotor spar concept include increased operational life and improved ballistic survivability at low life cycle cost.

  3. Design of radiation shields in nuclear reactor core

    International Nuclear Information System (INIS)

    Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

    2008-01-01

    This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

  4. Novel Natural Convection Heat Sink Design Concepts From First Principles

    Science.gov (United States)

    2016-06-01

    CONVECTION HEAT SINK DESIGN CONCEPTS FROM FIRST PRINCIPLES by Derek E. Fletcher June 2016 Thesis Advisor: Garth Hobson Second Reader...COVERED Master’s Thesis 4. TITLE AND SUBTITLE NOVEL NATURAL CONVECTION HEAT SINK DESIGN CONCEPTS FROM FIRST PRINCIPLES 5. FUNDING NUMBERS 6...CONVECTION HEAT SINK DESIGN CONCEPTS FROM FIRST PRINCIPLES Derek E. Fletcher Lieutenant Commander, United States Navy B.S., Southwestern

  5. Living closer to the environment: Housing design concept

    Directory of Open Access Journals (Sweden)

    Kosorić Vesna

    2011-01-01

    Full Text Available The main idea of this design concept is to strengthen the relationship and understanding between a man - resident and his environment. Residents are separated from the outdoor environment by glazing, which enables constant observation of environment from nearly all points of indoor space, encouraging positive feelings towards external world and understanding of the fragility of biosphere. Care for the environment should become a part of a man's nature and way of living, and it is the people who are expected to become the driving force of positive global changes towards sustainable development. The semisphere-like single family house of 14m in diameter has a multifunctional, multi-layer 'active' facade envelope. The envelope ensures constant visual contact of residents with the whole surroundings, while still providing comfort. The living space of the house reflects natural shapes which are organic rather than rectangular. Such indoor space becomes a part of the environment, rather than being protected, distanced and isolated from it. The house is designed to use solar energy 'passively' by absorption through insulated glazed envelope and 'actively' by outer skin layer on the first floor, made of stripes of flat semi-transparent polycrystalline photovoltaic (PV panels. In addition to its constructive role, the concrete core of the house acts as thermal mass and enables absorption and accumulation of thermal energy. The developed housing concept is applicable in different urban-design units and sets.

  6. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  7. Forming Core Elements for Strategic Design Management

    DEFF Research Database (Denmark)

    Beim, Anne; Vibæk, Kasper Sánchez

    2007-01-01

    . This evolution also leads to a demand for precise definitions of the values and qualities that can be used as managing tools in common building practice and it puts the traditional architectural design process under pressure. This paper outlines an approach to architectural quality as dealt with in the design...... and how the architectural potentials are realized when dealing with modern industrial processes are examined. To analyse and structure the empirical data, a model was developed consisting of four approaches for action. The approaches are categorized along different dichotomies in order to point out...... the model was tested in the architectural education at the Royal Danish Academy of Fine Arts - School of Architecture. The overall research project has two aims - to help offices identify the characteristics and specific methods of working with architectural quality in an industrialized context...

  8. A Minimum Shuffle Core Design Strategy for ESBWR

    International Nuclear Information System (INIS)

    Karve, A.A.; Fawcett, R.M.

    2008-01-01

    The Economic Simplified Boiling Water Reactor (ESBWR) is GEH's next evolution of advanced BWR technology. There are 1132 fuel bundles in the core and the thermal power is 4500 MWt. Similar to conventional plants there is an outage after a specified period of operation, when the plant shuts down. During the outage a specified fraction of fuel bundles are discharged from the core, it is loaded with the same fraction of fresh fuel, and fuel is shuffled to obtain an optimum core design that meets the goals for a successful operation of the next cycle. The discharge, load, and the associated shuffles are time-consuming and expensive tasks that impact the overall outage schedule and costs. Therefore, there is an incentive to keep maneuvers to a minimum and to perform them more efficiently. The benefits for a large core, such as the ESBWR with 1132 fuel bundles, are escalated. This study focuses on a core reload design strategy to minimize the total number of shuffles during an outage. A traditional equilibrium cycle is used as a reference basis, which sets the reference number of shuffles. In the minimum shuffle core design however, a set of two equilibrium cycles (N and N+1, referred to as a 'bi- equilibrium' cycle) is envisioned where the fresh fuel of cycle N (that becomes the once-burnt fuel of cycle N+1) ideally does not move in the two cycles. The cost of fuel efficiency is determined for obtaining such a core loading by comparing it to the traditional equilibrium cycle. There are several additional degrees of freedom when designing a bi-equilibrium cycle that could be utilized, and the potential benefits of these flexibilities are assessed. In summary, the feasibility of a minimum shuffle fuel cycle and core design for an ESBWR is studied. The cost of fuel efficiency is assessed in comparison to the traditional design. (authors)

  9. Reverse depletion method for PWR core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.; Kim, Y.J.

    1985-01-01

    Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. Prospects for even greater use of in-board fresh fuel loading are good as utilities seek to reduce core vessel fluence, mitigate pressurized thermal shock concerns, and extend vessel lifetime. Consequently, large numbers of burnable poison (BP) pins are being used to control the power peaking at the in-board fresh fuel positions. This has presented an additional complexity to the core reload design problem. In addition to determining the best location of each assembly in the core, the designer must concurrently determine the distribution of BP pins in the fresh fuel. A procedure was developed that utilizes the well-known Haling depletion to achieve an end-of-cycle (EOC) core state where the assembly pattern is configured in the absence of all control poison. This effectively separates the assembly assignment and BP distribution problems. Once an acceptable pattern at EOC is configured, the burnable and soluble poison required to control the power and core excess reactivity are solved for as unknown variables while depleting the cycle in reverse from the EOC exposure distribution to the beginning of cycle. The methods developed were implemented in an approved light water reactor licensing code to ensure the validity of the results obtained and provide for the maximum utility to PWR core reload design

  10. Concept design of the CFETR central solenoid

    International Nuclear Information System (INIS)

    Zheng, Jinxing; Song, Yuntao; Liu, Xufeng; Li, Jiangang; Wan, Yuanxi; Wan, Baonian; Ye, Minyou; Wu, Huan

    2015-01-01

    Highlights: • Main concept design work including coil's geometry, superconductor and support structure has been carried out. • The maximum magnetic field of CS coil is 11.9 T which is calculated by the coils’ operation current based on plasma equilibrium configuration. • The stray field in plasma area is less than 20 Gs under the CS coils’ operation currents designed for the plasma-heating phase. - Abstract: China Fusion Engineering Test Reactor (CFETR) superconducting tokamak is a national scientific research project of China with major and minor radius is 5.7 m and 1.6 m respectively. The magnetic field at the center of plasma with radius as R = 5.7 m is set to be 5.0 T. The major objective of the project is to build a fusion engineering tokamak reactor with fusion power in the range of 50–200 MW and should be self-sufficient by blanket. Six central solenoid coils of CFETR with same structure are made of Nb 3 Sn superconductor. Besides, the stray field in plasma area should be less than 20 Gs with the operation current of CS coils for plasma heating phase. The maximum magnetic field of CS coil is 11.9 T. It is calculated by the coils’ operation current based on plasma equilibrium configuration. The central solenoid needs to have enough stability margin under the condition of high magnetic field and strain. This paper discusses the design parameters, electromagnetic distribution, structure and stability analysis of the CS superconducting magnet for CFETR

  11. Design concepts of nuclear desalination plants

    International Nuclear Information System (INIS)

    2002-11-01

    Interest in using nuclear energy for producing potable water has been growing worldwide in the past decade. This has been motivated by a variety of factors, including economic competitiveness of nuclear energy, the growing need for worldwide energy supply diversification, the need to conserve limited supplies of fossil fuels, protecting the environment from greenhouse gas emissions, and potentially advantageous spin-off effects of nuclear technology for industrial development. Various studies, and at least one demonstration project, have been considered by Member States with the aim of assessing the feasibility of using nuclear energy for desalination applications under specific conditions. In order to facilitate information exchange on the subject area, the IAEA has been active for a number of years in compiling related technical publications. In 1999, an inter regional technical co-operation project on Integrated Nuclear Power and desalination System Design was launched to facilitate international collaboration for the joint development by technology holders and potential end users of an integrated nuclear desalination system. This publication presents material on the current status of nuclear desalination activities and preliminary design concepts of nuclear desalination plants, as made available to the IAEA by various Member States. It is aimed at planners, designers and potential end-users in those Member States interested in further assessment of nuclear desalination. Interested readers are also referred to two related and recent IAEA publications, which contain useful information in this area: Introduction of Nuclear Desalination: A Guidebook, Technical Report Series No. 400 (2000) and Safety Aspects of Nuclear Plants Coupled with Seawater Desalination Units, IAEA-TECDOC-1235 (2001)

  12. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  13. Status of the Astrid core at the end of the pre-conceptual design phase 1

    International Nuclear Information System (INIS)

    Chenaud, Ms.; Devictor, N.; Mignot, G.; Varaine, F.; Venard, C.; Martin, L.; Phelip, M.; Lorenzo, D.; Serre, F.; Bertrand, F.; Alpy, N.; Le-Flem, M.; Gavoille, P.; Lavastre, R.; Richard, P.; Verrier, D.; Schmitt, D.

    2013-01-01

    Within the framework of the ASTRID project, core design studies are being conducted by the CEA with support from AREVA and EDF. The pre-conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves limiting the consequences of 1) a hypothetical control rod withdrawal accident (by minimizing the core reactivity loss during the irradiation cycle), and 2) an hypothetical loss-of-flow accident (by reducing the sodium void worth). Two types of cores are being studied for the ASTRID project. The first is based on a 'large pin/small spacing wire' concept derived from the SFR V2b, while the other is based on an innovative CFV design. A distinctive feature of the CFV core is its negative sodium void worth. In 2011, the evaluation of a preliminary version (v1) of this CFV core for ASTRID underlined its potential capacity to improve the prevention of severe accidents. An improved version of the ASTRID CFV core (v2) was proposed in 2012 to comply with all the control rod withdrawal criteria, while increasing safety margins for all unprotected-loss-of-flow (ULOF) transients and improving the general design. This paper describes the CFV v2 design options and reports on the progress of the studies at the end of pre-conceptual design phase 1 concerning: - Core performance, - Intrinsic behavior during unprotected transients, - Simulation of severe accident scenarios, - Qualification requirements. The paper also specifies the open options for the materials, sub-assemblies, absorbers, and core monitoring that will continue to be studied during the conceptual design phase. (authors)

  14. A Core Design Approach Aimed at Sustainability and Intrinsic Safety

    International Nuclear Information System (INIS)

    Grasso, Giacomo

    2013-01-01

    The comprehensive approach adopted for the core design of all LFRs investigated within the LEADER project, proved to effectively drive the design to the fulfillment of the aimed sustainability performances, and the respect of the design constraints for the robust implementation of the inherent safety principle: • the ELFR core is able to operate adiabatically, with a very narrow reactivity swing along a 2.5 y cycle; • wide margins are provided for protecting the fuel and the structures even in case of unprotected transients, allowing for very long grace times

  15. BioCore Guide: A Tool for Interpreting the Core Concepts of Vision and Change for Biology Majors.

    Science.gov (United States)

    Brownell, Sara E; Freeman, Scott; Wenderoth, Mary Pat; Crowe, Alison J

    2014-01-01

    Vision and Change in Undergraduate Biology Education outlined five core concepts intended to guide undergraduate biology education: 1) evolution; 2) structure and function; 3) information flow, exchange, and storage; 4) pathways and transformations of energy and matter; and 5) systems. We have taken these general recommendations and created a Vision and Change BioCore Guide-a set of general principles and specific statements that expand upon the core concepts, creating a framework that biology departments can use to align with the goals of Vision and Change. We used a grassroots approach to generate the BioCore Guide, beginning with faculty ideas as the basis for an iterative process that incorporated feedback from more than 240 biologists and biology educators at a diverse range of academic institutions throughout the United States. The final validation step in this process demonstrated strong national consensus, with more than 90% of respondents agreeing with the importance and scientific accuracy of the statements. It is our hope that the BioCore Guide will serve as an agent of change for biology departments as we move toward transforming undergraduate biology education. © 2014 S. E. Brownell et al. CBE—Life Sciences Education © 2014 The American Society for Cell Biology. This article is distributed by The American Society for Cell Biology under license from the author(s). It is available to the public under an Attribution–Noncommercial–Share Alike 3.0 Unported Creative Commons License (http://creativecommons.org/licenses/by-nc-sa/3.0).

  16. Feasibility Study of Core Design with a Monte Carlo Code for APR1400 Initial core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jinsun; Chang, Do Ik; Seong, Kibong [KEPCO NF, Daejeon (Korea, Republic of)

    2014-10-15

    The Monte Carlo calculation becomes more popular and useful nowadays due to the rapid progress in computing power and parallel calculation techniques. There have been many attempts to analyze a commercial core by Monte Carlo transport code using the enhanced computer capability, recently. In this paper, Monte Carlo calculation of APR1400 initial core has been performed and the results are compared with the calculation results of conventional deterministic code to find out the feasibility of core design using Monte Carlo code. SERPENT, a 3D continuous-energy Monte Carlo reactor physics burnup calculation code is used for this purpose and the KARMA-ASTRA code system, which is used for a deterministic code of comparison. The preliminary investigation for the feasibility of commercial core design with Monte Carlo code was performed in this study. Simplified core geometry modeling was performed for the reactor core surroundings and reactor coolant model is based on two region model. The reactivity difference at HZP ARO condition between Monte Carlo code and the deterministic code is consistent with each other and the reactivity difference during the depletion could be reduced by adopting the realistic moderator temperature. The reactivity difference calculated at HFP, BOC, ARO equilibrium condition was 180 ±9 pcm, with axial moderator temperature of a deterministic code. The computing time will be a significant burden at this time for the application of Monte Carlo code to the commercial core design even with the application of parallel computing because numerous core simulations are required for actual loading pattern search. One of the remedy will be a combination of Monte Carlo code and the deterministic code to generate the physics data. The comparison of physics parameters with sophisticated moderator temperature modeling and depletion will be performed for a further study.

  17. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    Melo, C.A. de.

    1982-12-01

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author) [pt

  18. Refractory metal component technology for in-core sensor design

    International Nuclear Information System (INIS)

    Cannon, C.P.

    1986-02-01

    Within recent years, an increasing concern over reactor safety has prompted tests that characterize reactor core environments during transient conditions. Such tests include the Loss-of-Fluid-Tests (Idaho National Engineering Lab (INEL)), Severe Fuel Damage Tests (INEL), Core Debris Rubble Tests (Sandia National Laboratories (SNL)), and similar tests performed by foreign nations. The in-core sensors for these tests require refractory metal components to be compatible with electrical insulator materials as well as materials comprising highly corrosive service mediums. This paper presents the refractory metal technology utilized to provide basic sensor designs in the above mentioned reactor tests

  19. Development of core thermal-hydraulics module for intelligent reactor design system (IRDS)

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki; Fujii, Sadao.

    1994-08-01

    We have developed an innovative reactor core thermal-hydraulics module where a designer can easily and efficiently evaluate his design concept of a new type reactor in the thermal-hydraulics field. The main purpose of this module is to decide a feasible range of basic design parameters of a reactor core in a conceptual design stage of a new type reactor. The module is to be implemented in Intelligent Reactor Design System (IRDS). The module has the following characteristics; 1) to deal with several reactor types, 2) four thermal hydraulics and fuel behavior analysis codes are installed to treat different type of reactors and design detail, 3) to follow flexibly modification of a reactor concept, 4) to provide analysis results in an understandable way so that a designer can easily evaluate feasibility of his concept, and so on. The module runs on an engineering workstation (EWS) and has a user-friendly man-machine interface on a pre- and post-processing. And it is equipped with a function to search a feasible range called as Design Window, for two design parameters by artificial intelligence (AI) technique and knowledge engineering. In this report, structure, guidance for users of an usage of the module and instruction of input data for analysis modules are presented. (author)

  20. A Concept Transformation Learning Model for Architectural Design Learning Process

    Science.gov (United States)

    Wu, Yun-Wu; Weng, Kuo-Hua; Young, Li-Ming

    2016-01-01

    Generally, in the foundation course of architectural design, much emphasis is placed on teaching of the basic design skills without focusing on teaching students to apply the basic design concepts in their architectural designs or promoting students' own creativity. Therefore, this study aims to propose a concept transformation learning model to…

  1. In-reactor testing of the closed cycle gas core reactor---the nuclear light bulb concept

    International Nuclear Information System (INIS)

    Gauntt, R.O.; Slutz, S.A.; Harms, G.A.; Latham, T.S.; Roman, W.C.; Rodgers, R.J.

    1993-01-01

    The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>1800 s) and thrust (>445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (∼4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept

  2. Characteristics of fast reactor core designs and closed fuel cycle

    International Nuclear Information System (INIS)

    Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N.

    2007-01-01

    On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

  3. Hardware concepts for a large low-energetics LMFBR core. Final report

    International Nuclear Information System (INIS)

    Hutter, E.; Batch, R.V.

    1980-12-01

    A design study was made to identify a practical set of hardware configurations that would embody the requirements developed in the numerical study of a low-energetics core and blanket for a prototype large breeder reactor. Dimensioned drawings are presented for fuel, blanket, reflector/shield, and control rod subassemblies. A horizontal cross section drawing shows how these subassemblies are arranged in the total core/blanket assembly. A core support is illustrated showing a dual plenums arrangement

  4. Core design of super LWR with double tube water rods

    International Nuclear Information System (INIS)

    Wu, Jianhui; Oka, Yoshiaki

    2014-01-01

    Highlights: • Supercritical light water cooled and moderated reactor with double tube water rods is developed. • Double-row fuel rod assembly and out-in fuel loading pattern are applied. • Separation plates in peripheral assemblies increase average outlet temperature. • Neutronic and thermal design criteria are satisfied during the cycle. - Abstract: Double tube water rods are employed in core design of super LWR to simplify the upper core structure and refueling procedure. The light water moderator flows up in the inner tube from the bottom of the core, then, changes the flow direction at the top of the core into the outer tube and flows out at the bottom of the core. It eliminates the moderator guide/distribution tubes into the single tube water rods from the top dome of the reactor pressure vessel of the previous super LWR design. Two rows of fuel rods are filled between the water rods in the fuel assembly. Out-in refueling pattern is adopted to flatten radial power distribution. The peripheral fuel assemblies of the core are divided into four flow zones by separation plates for increasing the average core outlet temperature. Three enrichment zones are used for axial power flattening. The equilibrium core is analyzed based on neutronic/thermal-hydraulic coupled model. The results show that, by applying the separation plates in peripheral fuel assemblies and low gadolinia enrichment, the maximum cladding surface temperature (MCST) is limited to 653 °C with the average outlet temperature of 500 °C. The inherent safety is satisfied by the negative void reactivity effects and sufficient shutdown margin

  5. Status of core nuclear design technology for future fuel

    International Nuclear Information System (INIS)

    Joo, Hyung Kook; Jung, Hyung Guk; Noh, Jae Man; Kim, Yeong Il; Kim, Taek Kyum; Gil, Choong Sup; Kim, Jung Do; Kim, Young Jin; Sohn, Dong Seong

    1997-01-01

    The effective utilization of nuclear resource is more important factor to be considered in the design of next generation PWR in addition to the epochal consideration on economics and safety. Assuming that MOX fuel can be considered as one of the future fuel corresponding to the above request, the establishment of basic technology for the MOX core design has been performed : : the specification of the technical problem through the preliminary core design and nuclear characteristic analysis of MOX, the development and verification of the neutron library for lattice code, and the acquisition of data to be used for verification of lattice and core analysis codes. The following further studies will be done in future: detailed verification of library E63LIB/A, development of the spectral history effect treatment module, extension of decay chain, development of new homogenization for the MOX fuel assembly. (author). 6 refs., 7 tabs., 2 figs

  6. The APR1400 Core Design by Using APA Code System

    International Nuclear Information System (INIS)

    Choi, Yu Sun; Koh, Byung Marn

    2008-01-01

    The nuclear design for APR1400 has been performed to prepare the core model for Automatic Load Follow Operation Simulation. APA (ALPHA/ PHOENIXP/ ANC) code system is a tool for the multi-cycle depletion calculations for APR1400. Its detail versions for ALPHA, PHOENIX-P and ANC are 8.9.3, 8.6.1 and 8.10.5, respectively. The first and equilibrium core depletion calculations for APR1400 have been performed to assure the target cycle length and confirm the safety parameters. The parameters are satisfied within limitation about nuclear design criteria. This APR1400 core models will be based on the design parameters for APR1400 Simulator

  7. Design and economic implications of heterogeneity in an LMFBR core

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1983-01-01

    Much emphasis is currently being placed in LMFBR design on reducing both the capital cost and the fuel cycle cost of an LMFBR to insure its economic competativeness without a rapid increase in the uranium prices. In this study the relationship between two core design options, their neutronic consequences, and their effect on fuel cycle cost are analyzed. The two design options are the selection of pin diameter and the degree of heterogeneity. In the case of a heterogeneous core, with a low sodium void reactivity worth this ratio of fertile internal blanket to driver assemblies is generally about 0.40. However, some advantages of cores with heterogeneity of 0.08 to 0.2 for a fixed pin diameter have been reported

  8. Mathematical concepts for mechanical engineering design

    CERN Document Server

    Asli, Kaveh Hariri; Aliyev, Soltan Ali Ogli

    2013-01-01

    PrefaceIntroductionHeat Flow: From Theory to PracticeDispersed Fluid and Ideal Fluid MechanicsModeling for Pressure Wave into Water PipelineHeat Transfer and Vapor BubbleMathematical Concepts and Computational Approaches on Hydrodynamics InstabilityMathematical Concepts and Dynamic ModelingModeling for Predictions of Air Entrance into Water PipelineIndex

  9. Core design methodology and software for Temelin NPP

    International Nuclear Information System (INIS)

    Havluj, F; Hejzlar, J.; Klouzal, J.; Stary, V.; Vocka, R.

    2011-01-01

    In the frame of the process of fuel vendor change at Temelin NPP in the Czech Republic, where, starting since 2010, TVEL TVSA-T fuel is loaded instead of Westinghouse VVANTAGE-6 fuel, new methodologies for core design and core reload safety evaluation have been developed. These documents are based on the methodologies delivered by TVEL within the fuel contract, and they were further adapted according to Temelin NPP operational needs and according to the current practice at NPP. Along with the methodology development the 3D core analysis code ANDREA, licensed for core reload safety evaluation in 2010, have been upgraded in order to optimize the safety evaluation process. New sequences of calculations were implemented in order to simplify the evaluation of different limiting parameters and output visualization tools were developed to make the verification process user friendly. Interfaces to the fuel performance code TRANSURANUS and sub-channel analysis code SUBCAL were developed as well. (authors)

  10. Stationary liquid fuel fast reactor SLFFR – Part I: Core design

    Energy Technology Data Exchange (ETDEWEB)

    Jing, T.; Yang, G.; Jung, Y.S.; Yang, W.S., E-mail: yang494@purdue.edu

    2016-12-15

    Highlights: • An innovative fast reactor concept SLFFR based on liquid metal fuel is proposed for TRU burning. • A compact core design of 1000 MWt SLFFR is developed to achieve a zero conversion ratio and passive safety. • The core size and the control requirement are significantly reduced compared to the conventional solid fuel reactor with same conversion ratio. - Abstract: For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named the stationary liquid fuel fast reactor (SLFFR) has been proposed based on a stationary molten metallic fuel. A compact core design of a 1000 MWt SLFFR has been developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches have been adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses have been performed to evaluate the steady-state performance characteristics. The analysis results indicate that the SLFFR of a zero TRU conversion ratio is feasible while satisfying the conservatively imposed thermal design constraints. A theoretical maximum TRU consumption rate of 1.01 kg/day is achieved with uranium-free fuel. Compared to the solid fuel reactors with the same TRU conversion ratio, the core size and the reactivity control requirement are reduced significantly. The primary and secondary control systems provide sufficient shutdown margins, and the calculated reactivity feedback coefficients show that the prompt fuel expansion coefficient is sufficiently negative.

  11. Advanced composites structural concepts and materials technologies for primary aircraft structures: Design/manufacturing concept assessment

    Science.gov (United States)

    Chu, Robert L.; Bayha, Tom D.; Davis, HU; Ingram, J. ED; Shukla, Jay G.

    1992-01-01

    Composite Wing and Fuselage Structural Design/Manufacturing Concepts have been developed and evaluated. Trade studies were performed to determine how well the concepts satisfy the program goals of 25 percent cost savings, 40 percent weight savings with aircraft resizing, and 50 percent part count reduction as compared to the aluminum Lockheed L-1011 baseline. The concepts developed using emerging technologies such as large scale resin transfer molding (RTM), automatic tow placed (ATP), braiding, out-of-autoclave and automated manufacturing processes for both thermoset and thermoplastic materials were evaluated for possible application in the design concepts. Trade studies were used to determine which concepts carry into the detailed design development subtask.

  12. Status of experimental data for the VHTR core design

    Energy Technology Data Exchange (ETDEWEB)

    Park, Won Seok; Chang, Jong Hwa; Park, Chang Kue

    2004-05-01

    The VHTR (Very High Temperature Reactor) is being emerged as a next generation nuclear reactor to demonstrate emission-free nuclear-assisted electricity and hydrogen production. The VHTR could be either a prismatic or pebble type helium cooled, graphite moderated reactor. The final decision will be made after the completion of the pre-conceptual design for each type. For the pre-conceptual design for both types, computational tools are being developed. Experimental data are required to validate the tools to be developed. Many experiments on the HTGR (High Temperature Gas-cooled Reactor) cores have been performed to confirm the design data and to validate the design tools. The applicability and availability of the existing experimental data have been investigated for the VHTR core design in this report.

  13. PGSFR Core Thermal Design Procedure to Evaluate the Safety Margin

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Kim, Sang-Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The Korea Atomic Energy Research Institute (KAERI) has performed a SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal design is to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damage in SFR subassemblies arises from a creep induced failure. The creep limit is evaluated based on the maximum cladding temperature, power, neutron flux, and uncertainties in the design parameters, as shown in Fig. 1. In this work, the core thermal design procedures are compared to verify the present PGSFR methodology based on the nuclear plant design criteria/guidelines and previous SFR thermal design methods. The PGSFR core thermal design procedure is verified based on the nuclear plant design criteria/guidelines and previous methods in LWRs and SFRs. The present method aims to directly evaluate the fuel cladding failure and to assure more safety margin. The 2 uncertainty is similar to 95% one-side tolerance limit of 1.96 in LWRs. The HCFs, ITDP, and MCM reveal similar uncertainty propagation for cladding midwall temperature for typical SFR conditions. The present HCFs are mainly employed from the CRBR except the fuel-related uncertainty such as an incorrect fuel distribution. Preliminary PGSFR specific HCFs will be developed by the end of 2015.

  14. Concept of object-oriented intelligent support for nuclear reactor designing

    International Nuclear Information System (INIS)

    Yoshikawa, H.; Gofuku, A.

    1991-01-01

    A concept of object-oriented intelligent CAD/CAE environment is proposed for the conceptual designing of advanced nuclear reactor system. It is composed of (i) object-oriented frame-structure database which represents the hierarchical relationship of the composite elements of reactor core and the physical properties, and (ii) object-oriented modularization of the elementary calculation processes, which are needed for reactor core design analysis. As an example practise, an object-oriented frame structure is constructed for representing a 3D configuration of a special fuel element of a space reactor design, by using a general-purpose expert system shell ESHELL/X. (author)

  15. Advances in architectural concepts to support distributed systems design

    NARCIS (Netherlands)

    Ferreira Pires, Luis; Vissers, C.A.; van Sinderen, Marten J.

    1993-01-01

    This paper presents and discusses some architectural concepts for distributed systems design. These concepts are derived from an analysis of limitations of some currently available standard design languages. We conclude that language design should be based upon the careful consideration of

  16. The workspace design concept: A new framework of participatory ergonomics

    OpenAIRE

    Broberg, Ole

    2007-01-01

    The concept of Workspace Design is presented as a potential new approach for ergonomists and consultants in the occupational health service. The concept is aimed as an intervention and facilitation strategy in the early stages of design processes leading to new workplaces. Preliminary results from a case study demonstrate how Workspace Design can contribute to a technical change process.

  17. Learning Gains for Core Concepts in a Serious Game on Scientific Reasoning

    Science.gov (United States)

    Forsyth, Carol; Pavlik, Philip, Jr.; Graesser, Arthur C.; Cai, Zhiqiang; Germany, Mae-lynn; Millis, Keith; Dolan, Robert P.; Butler, Heather; Halpern, Diane

    2012-01-01

    "OperationARIES!" is an Intelligent Tutoring System that teaches scientific inquiry skills in a game-like atmosphere. Students complete three different training modules, each with natural language conversations, in order to acquire deep-level knowledge of 21 core concepts of research methodology (e.g., correlation does not mean…

  18. Validating a Conceptual Framework for the Core Concept of "Cell-Cell Communication"

    Science.gov (United States)

    Michael, Joel; Martinkova, Patricia; McFarland, Jenny; Wright, Ann; Cliff, William; Modell, Harold; Wenderoth, Mary Pat

    2017-01-01

    We have created and validated a conceptual framework for the core physiology concept of "cell-cell communication." The conceptual framework is composed of 51 items arranged in a hierarchy that is, in some instances, four levels deep. We have validated it with input from faculty who teach at a wide variety of institutional types. All…

  19. Validating a Conceptual Framework for the Core Concept of ”Cell-cell Communication”

    Czech Academy of Sciences Publication Activity Database

    Michael, J.; Martinková, Patrícia; McFarland, J.L.; Wright, A.; Cliff, W.; Modell, H.; Wenderoth, M.P.

    2017-01-01

    Roč. 41, č. 2 (2017), s. 260-265 ISSN 1043-4046 R&D Projects: GA ČR GJ15-15856Y Institutional support: RVO:67985807 Keywords : conceptual framework * core concept * cell-cell communication * physiology Subject RIV: AM - Education OBOR OECD: Education , general; including training, pedagogy, didactics [and education systems] Impact factor: 1.755, year: 2016

  20. Core designs for the de-regulated market

    International Nuclear Information System (INIS)

    Almberger, J.; Bernro, R.; Pettersson, H.

    1999-01-01

    Complete text of publication follows: The electricity market deregulation in the Nordic countries encourages innovations and cost reductions for power production in the Vattenfall reactors. The competition on the electricity market is strong, electricity price reductions dramatic and uncertainties about the future power demand is large. In the fuel area this situation has given increased attention to traditional areas like flexibility in power production, improved core designs, need for margins (improved fuel designs), improved surveillance, decreased lead times. At Vattenfall new fuel designs are already being implemented following the last fuel purchase, for which flexibility and margins, were given high values in the evaluations with the multipurpose task of eliminating fuel related problems and meeting the future market situation. This strategy has given Vattenfall a flying start to meeting the demands of the de-regulated market. What has been added are broad studies undertaken to investigate the various route into the future with respect to finding the most effective strategies for fuel and core design and optimization. In the present paper the Vattenfall priorities for fuel designs and margins are presented in a schematic manner summarizing the results of the last fuel purchase and also presenting the current program for LFAs. Technical limitations, licensing and R and D aspects, with respect to improving the fuel utilization will be mentioned. The main focus in the paper is on the broad study carried out in the PWR core design area. Driven by the relatively low power demand various possibilities for higher production flexibility have been investigated specifically extended coast-down, coast-up and yearly load follow. Further to reduce the costs for fuel consumption improvements in core designs have been studied: improved low leakage loading patterns, low enriched end zones, improved Gd designs etc. Main results and conclusions of the core design studies will

  1. High Performance Systolic Array Core Architecture Design for DNA Sequencer

    Directory of Open Access Journals (Sweden)

    Saiful Nurdin Dayana

    2018-01-01

    Full Text Available This paper presents a high performance systolic array (SA core architecture design for Deoxyribonucleic Acid (DNA sequencer. The core implements the affine gap penalty score Smith-Waterman (SW algorithm. This time-consuming local alignment algorithm guarantees optimal alignment between DNA sequences, but it requires quadratic computation time when performed on standard desktop computers. The use of linear SA decreases the time complexity from quadratic to linear. In addition, with the exponential growth of DNA databases, the SA architecture is used to overcome the timing issue. In this work, the SW algorithm has been captured using Verilog Hardware Description Language (HDL and simulated using Xilinx ISIM simulator. The proposed design has been implemented in Xilinx Virtex -6 Field Programmable Gate Array (FPGA and improved in the core area by 90% reduction.

  2. Neutronic design of mixed oxide-silicide cores for the core conversion of rsg-gas reactor

    International Nuclear Information System (INIS)

    Sembiring, Tagor Malem; Tukiran; Pinem surian; Febrianto

    2001-01-01

    The core conversion of rsg-gas reactor from an all-oxide (U 3 O 8 -Al) core, through a series of mixed oxide-silicide core, to an all-silicide (U 3 Si 2 -Al) core for the same meat density of 2.96 g U/cc is in progress. The conversion is first step of the step-wise conversion and will be followed by the second step that is the core conversion from low meat density of silicide core, through a series of mixed lower-higher density of silicide core, to an all-higher meat density of 3.55 g/cc core. Therefore, the objectives of this work is to design the mixed cores on the neutronic performance to achieve safety a first full-silicide core for the reactor with the low uranium meat density of 2.96gU/cc. The neutronic design of the mixed cores was performed by means of Batan-EQUIL-2D and Batan-3DIFF computer codes for 2 and 3 dimension diffusion calculation, respectively. The result shows that all mixed oxide-silicide cores will be feasible to achieve safety a fist full-silicide core. The core performs the same neutronic core parameters as those of the equilibrium silicide core. Therefore, the reactor availability and utilization during the core conversion is not changed

  3. Application of green concept in mechanical design and manufacture

    Science.gov (United States)

    Liu, Xing ping

    2017-11-01

    With the development of productive forces, the relationship between human and nature is becoming tight increasingly, especially environmental pollution and resource consumption that comes from equipment manufacturing industry mainly. Green development concept is a new concept which can solve the current ecological environment. The philosophical foundation and theoretical basis of green idea are expounded through the study of scientific development and green concept. The difference between the traditional design and the green design is analyzed; the meaning and content of the mechanical design for green concept are discussed. And the evaluation method of green design is discussed too. The significance of green development concept in the mechanical design and manufacturing science is pinpointed clearly. The results show that the implementation of green design under the mechanical design, from the source of pollution control to achieve green manufacturing, is the only way to achieve sustainable development.

  4. New design on air-core resistive NMR imaging magnet

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Yan; Mingwu, Fan; Yixin, Miao

    1984-08-01

    A new type of NMR imaging air-core resistive magnet is designed. Based on the BIM Magnetostatic calculation the resultant four equiradial coils structure with optimized shapes of cross section possesses a larger spherical working volume obviously, comparing with the common four-coils imaging magnet. The manufacturing tolerance is also calculated.

  5. 300 MWe Burner Core Design with two Enrichment Zoning

    International Nuclear Information System (INIS)

    Song, Hoon; Kim, Sang Ji; Kim, Yeong Il

    2008-01-01

    KAERI has been developing the KALIMER-600 core design with a breakeven fissile conversion ratio. The core is loaded with a ternary metallic fuel (TRU-U-10Zr), and the breakeven characteristics are achieved without any blanket assembly. As an alternative plan, a KALIMER-600 burner core design has been also performed. In the early stage of the development of a fast reactor, the main purpose is an economical use of a uranium resource but nowadays in addition to the maximum utilization of a uranium resource, the burning of a high level radioactive waste is taken as an additional interest for the harmony of the environment. In way of constructing the commercial size reactor which has the power level ranging from 800 MWe to 1600 MWe, the demonstration reactor which has the power level ranging from 200 MWe to 600 MWe was usually constructed for the midterm stage to commercial size reactor. In this paper, a 300 MWe burner core design was performed with purpose of demonstration reactor for KALIMER-600 burner of 600 MWe. As a means to flatten the power distribution, instead of a single fuel enrichment scheme adapted in design of KALIMER-600 burner, the 2 enrichment zoning approach was adapted

  6. An integral reactor design concept for a nuclear co-generation plant

    International Nuclear Information System (INIS)

    Lee, D.J.; Kim, J.I.; Kim, K.K.; Chang, M.H.; Moon, K.S.

    1997-01-01

    An integral reactor concept for nuclear cogeneration plant is being developed at KAERI as an attempt to expand the peaceful utilization of well established commercial nuclear technology, and related industrial infrastructure such as desalination technology in Korea. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway to evaluate the characteristics of various passive safety concepts and provide the proper technical data for the conceptual design. This paper describes the preliminary safety and design concepts of the advanced integral reactor. Salient features of the design are hexagonal core geometry, once-through helical steam generator, self-pressurizer, and seismic resistant fine control CEDMS, passive residual heat removal system, steam injector driven passive containment cooling system. (author)

  7. KATS experiments to simulate corium spreading in the EPR core catcher concept

    International Nuclear Information System (INIS)

    Eppinger, B.; Fieg, G.; Schuetz, W.; Stegmaier, U.

    2001-01-01

    In future Light Water Reactors special devices (core catchers) might be required to prevent containment failure by basement erosion after reactor pressure vessel melt-through during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher de-vices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent cooling by flooding with water. Therefore a series of experiments to investigate high temperature melt spreading on flat surfaces has been carried out using alumina-iron thermite melts as a simulant. The oxidic thermite melt is conditioned by adding other oxides to simulate a realistic corium melt as close as possible. Spreading of oxidic and metallic melts have been performed in one- and two-dimensional geometry. Substrates were chemically inert ceramic layers, dry concrete and concrete with a shallow water layer on top. (authors)

  8. Capturing design knowledge in formal concept definitions

    NARCIS (Netherlands)

    Leeuwen, van J.P.; Vries, de B.; Timmermans, H.J.P.; Vries, de B.

    2000-01-01

    For support of creativity in architectural design, design systems must be provided with information models that are flexible enough to follow the dynamic way of designers in handling early design information. This paper discusses a framework for information modelling using Features that answers this

  9. Westinghouse loading pattern search methodology for complex core designs

    International Nuclear Information System (INIS)

    Chao, Y.A.; Alsop, B.H.; Johansen, B.J.; Morita, T.

    1991-01-01

    Pressurized water reactor core designs have become more complex and must meet a plethora of design constraints. Trends have been toward longer cycles with increased discharge burnup, increased burnable absorber (BA) number, mixed BA types, reduced radial leakage, axially blanketed fuel, and multiple-batch feed fuel regions. Obtaining economical reload core loading patterns (LPs) that meet design criteria is a difficult task to do manually. Automated LP search tools are needed. An LP search tool cannot possibly perform an exhaustive search because of the sheer size of the combinatorial problem. On the other hand, evolving complexity of the design features and constraints often invalidates expert rules based on past design experiences. Westinghouse has developed a sophisticated loading pattern search methodology. This methodology is embodied in the LPOP code, which Westinghouse nuclear designers use extensively. The LPOP code generates a variety of LPs meeting design constraints and performs a two-cycle economic evaluation of the generated LPs. The designer selects the most appropriate patterns for fine tuning and evaluation by the design codes. This paper describes the major features of the LPOP methodology that are relevant to fulfilling the aforementioned requirements. Data and examples are also provided to demonstrate the performance of LPOP in meeting the complex design needs

  10. ASTRID core: Design objectives, design approach, and R&D in support

    International Nuclear Information System (INIS)

    Mignot, G.; Devictor, N.

    2012-01-01

    ASTRID core design is mainly guided by safety objectives: 1. Prevention of the core meltdown accident: To prevent meltdown accidents: - by a natural behavior of the core and the reactor (no actuation of the two shutdown systems); - with adding passive complementary systems if natural behavior is not sufficient for some transient cases. 2. Mitigation of the fusion accident: To garantee that core fusion accidents don’t lead to significant mechanical energy release, whatever initiator event: - by a natural core behavior; - with adding specific mitigation dispositions in case of natural behavior is not suffficient

  11. Concept Generation for Design Creativity A Systematized Theory and Methodology

    CERN Document Server

    Taura, Toshiharu

    2013-01-01

    The concept generation process seems like an intuitional thought: difficult to capture and perform, although everyone is capable of it. It is not an analytical process but a synthetic process which has yet to be clarified. Furthermore, new research methods for investigating the concept generation process—a very difficult task since the concept generation process is driven by inner feelings deeply etched in the mind—are necessary to establish its theory and methodology.  Concept Generation for Design Creativity—A Systematized Theory and Methodology presents the concept generation process both theoretically and methodologically. Theoretically, the concept generation process is discussed by comparing metaphor, abduction, and General Design Theory from the perspective of similarities and dissimilarities. Property mapping, concept blending, and concept integration in thematic relation have been explained methodologically. So far, these theories and methods have been discussed independently, and the relation...

  12. GCRA review and appraisal of HTGR reactor-core-design program

    International Nuclear Information System (INIS)

    1980-09-01

    The reactor-core-design program has as its principal objective and responsibility the design and resolution of major technical issues for the reactor core and core components on a schedule consistent with the plant licensing and construction program. The task covered in this review includes three major design areas: core physics, core thermal and hydraulic performance fuel element design, and in-core fuel performance evaluation

  13. Development of out-of-core concepts for a supercritical-water, pressure-tube reactor

    International Nuclear Information System (INIS)

    Diamond, W.T.

    2010-01-01

    One of the Generation IV programs at Chalk River Laboratories has as its prime focus the development of out-of-core concepts for the SuperCritical Water (SCW) pressure tube reactor under development in Canada. A number of technical issues associated with the interface of out-of-core components and the pressure tubes of a SCW pressure tube reactor are being investigated. This article focuses on several aspects of out-of-core components and layouts, building upon concepts that have been developed during the past few years. The efforts are strongly focused on concepts for a fuel channel that can be fabricated with the tight lattice pitch (typically 230 to 250 mm) that may be required for some applications such as utilization of a thorium fuel cycle. It is not practical to adapt concepts with a tight lattice pitch while using the thicker materials required for the higher temperatures and pressures required for supercritical operation. A change in lattice pitch or configuration is required to accommodate the component size increases. This presentation will cover a number of new concepts developed to produce feeders and end fittings for the harsh conditions of a SCW pressure tube reactor. These components are then developed into conceptual models of a Gen IV pressure tube reactor mounted in both horizontal and vertical orientations. Full 3-D solid models of both concepts will be demonstrated as well as a 1/10th-scale model of one face of a horizontal concept that has been built from components made with a 3-D printer. (author)

  14. Mechatronical Aided Concept (MAC) in Intelligent Transport Vehicles Design

    OpenAIRE

    Pavel Pavlasek

    2003-01-01

    This article deals with the principles of synergy effect of mechatronical aided concept (MAC) to the design of intelligent transport vehicles products applying CA technologies and virtual reality design methods. Also includes presentation of intelligent railway vehicle development.

  15. Project Design Concept for Monitoring and Control System

    International Nuclear Information System (INIS)

    MCGREW, D.L.

    2000-01-01

    This Project Design Concept represents operational requirements established for use in design the tank farm Monitoring and Control System. These upgrades are included within the scope of Project W-314, Tank Farm Restoration and Safe Operations

  16. Intelligent system for conceptural design of new reactor cores

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki

    1995-01-01

    The software system IRDS has been developed at Japan Atomic Energy Research Institute to support the conceptual design of a new type of reactor core in the fields of neutronics, thermohydraulics, and fuel behavior. IRDS involves various analysis codes, database, and man-machine interfaces that efficiently support a whole design process on a computer. The main purpose of conceptual design is to decide an optimal set of basic design parameters. Designers usually carry out many parametric survey calculations and search a design window (DW), which is a feasible parameter range satisfying design criteria and goals. An automatic DW search function is installed to support such works. The man-machine interface based on menu windows will enable nonspecialists to use various analysis codes easily

  17. Adaption of the PARCS Code for Core Design Audit Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyong Chol; Lee, Young Jin; Uhm, Jae Beop; Kim, Hyunjik [Nuclear Safety Evaluation, Daejeon (Korea, Republic of); Jeong, Hun Young; Ahn, Seunghoon; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    The eigenvalue calculation also includes quasi-static core depletion analyses. PARCS has implemented variety of features and has been qualified as a regulatory audit code in conjunction with other NRC thermal-hydraulic codes such as TRACE or RELAP5. In this study, as an adaptation effort for audit applications, PARCS is applied for an audit analysis of a reload core design. The lattice physics code HELIOS is used for cross section generation. PARCS-HELIOS code system has been established as a core analysis tool. Calculation results have been compared on a wide spectrum of calculations such as power distribution, critical soluble boron concentration, and rod worth. A reasonable agreement between the audit calculation and the reference results has been found.

  18. Development of the Biological Experimental Design Concept Inventory (BEDCI)

    Science.gov (United States)

    Deane, Thomas; Nomme, Kathy; Jeffery, Erica; Pollock, Carol; Birol, Gulnur

    2014-01-01

    Interest in student conception of experimentation inspired the development of a fully validated 14-question inventory on experimental design in biology (BEDCI) by following established best practices in concept inventory (CI) design. This CI can be used to diagnose specific examples of non-expert-like thinking in students and to evaluate the…

  19. Advanced Concept Architecture Design and Integrated Analysis (ACADIA)

    Science.gov (United States)

    2017-11-03

    1 Advanced Concept Architecture Design and Integrated Analysis (ACADIA) Submitted to the National Institute of Aerospace (NIA) on...Research Report 20161001 - 20161030 Advanced Concept Architecture Design and Integrated Analysis (ACADIA) W911NF-16-2-0229 8504Cedric Justin, Youngjun

  20. The workspace design concept: A new framework of participatory ergonomics

    DEFF Research Database (Denmark)

    Broberg, Ole

    2007-01-01

    The concept of Workspace Design is presented as a potential new approach for ergonomists and consultants in the occupational health service. The concept is aimed as an intervention and facilitation strategy in the early stages of design processes leading to new workplaces. Preliminary results fro...

  1. Business Service Outsourcing: An Evolution of Concentration on Core Business Concepts and Transaction Cost Economies

    Directory of Open Access Journals (Sweden)

    Ahmad H. Juma'h

    2000-08-01

    Full Text Available Present and evaluate the literature with respect to business service outsourcing (BSO, with a particular concentration on a common form of BSO, information technology (IT. The review defines the issues with respect to BSO in terms of drivers and motivations, as well as internal and external implications for BSO companies and their contractors. BSO is an evolution of transaction cost theory and concentration on core business concepts. Although, there are several attempts to explain when a company should outsource, these theoretical frameworks are difficult to apply in practice since satisfaction is a function of expectation and the identification of activities are core commodity is not straightforward.

  2. Design and performance study of the helium-cooled T-tube divertor concept

    International Nuclear Information System (INIS)

    Ihli, T.; Raffray, A.R.; Abdel-Khalik, S.I.; Shin, S.

    2007-01-01

    The ARIES-CS study has been launched with the goal of developing through physics and engineering optimization an attractive power plant concept based on a compact stellarator configuration. The study included an effort to characterize the divertor location and corresponding heat load distribution, and to develop a He-cooled divertor concept that could accommodate a heat flux of at least 10 MW/m 2 , and that would integrate well with the other power core components. This paper describes the design study of this divertor concept, which, although developed for a compact stellarator, is well suited for a tokamak configuration also

  3. Improving Battery Reactor Core Design Using Optimization Method

    International Nuclear Information System (INIS)

    Son, Hyung M.; Suh, Kune Y.

    2011-01-01

    The Battery Omnibus Reactor Integral System (BORIS) is a small modular fast reactor being designed at Seoul National University to satisfy various energy demands, to maintain inherent safety by liquid-metal coolant lead for natural circulation heat transport, and to improve power conversion efficiency with the Modular Optimal Balance Integral System (MOBIS) using the supercritical carbon dioxide as working fluid. This study is focused on developing the Neutronics Optimized Reactor Analysis (NORA) method that can quickly generate conceptual design of a battery reactor core by means of first principle calculations, which is part of the optimization process for reactor assembly design of BORIS

  4. Design and analysis of PCRV core cavity closure

    International Nuclear Information System (INIS)

    Lee, T.T.; Schwartz, A.A.; Koopman, D.C.A.

    1980-05-01

    Design requirements and considerations for a core cavity closure which led to the choice of a concrete closure with a toggle hold-down as the design for the Gas-Cooled Fast Breeder Reactor (GCFR) plant are discussed. A procedure for preliminary stress analysis of the closure by means of a three-dimensional finite element method is described. A limited parametric study using this procedure indicates the adequacy of the present closure design and the significance of radial compression developed as a result of inclined support reaction

  5. Overview of PEC core design and requirements for PEC core restraint systems

    International Nuclear Information System (INIS)

    Cecchini, F.

    1984-01-01

    The Italian PEC reactor is an experimental loop type fast reactor of 120 MW thermal. Its main purpose is the in-pile development of fast reactor fuel. The mechanical principles in PEC core design and current modifications to ensure a safe seismic perturbation and shutdown are discussed in this paper. These anti-seismic modifications are aimed to limit the extent of reactivity perturbation during the seismic event and to guarantee control rod entry at any time during the seismic event

  6. Application Design Library With gamification concept

    Directory of Open Access Journals (Sweden)

    Nisaul Barokati

    2017-03-01

    Full Text Available The library is an effort to maintain and improve the efficiency and effectiveness of the learning process. Various means have been used to enhance the library's role in community development. One way is to develop a concept and a different orientation in the management system or the library. The concept in question is gamification. Gamification is a process with the aim of changing jobs or activities that usually go tedious and less enjoyable to be more interesting and fun to do. One approach taken is to reward both virtual and non-virtual that can increase people's motivation to do something, in this case, is to visit and take advantage of the functions and library facilities. This research resulted in a model that features a library application with the concept of gamification. The model can be implemented into an application that will increase visits and activities at the library. Thus the function and the main purpose of the library be met..

  7. Designing the colorectal cancer core dataset in Iran

    Directory of Open Access Journals (Sweden)

    Sara Dorri

    2017-01-01

    Full Text Available Background: There is no need to explain the importance of collection, recording and analyzing the information of disease in any health organization. In this regard, systematic design of standard data sets can be helpful to record uniform and consistent information. It can create interoperability between health care systems. The main purpose of this study was design the core dataset to record colorectal cancer information in Iran. Methods: For the design of the colorectal cancer core data set, a combination of literature review and expert consensus were used. In the first phase, the draft of the data set was designed based on colorectal cancer literature review and comparative studies. Then, in the second phase, this data set was evaluated by experts from different discipline such as medical informatics, oncology and surgery. Their comments and opinion were taken. In the third phase refined data set, was evaluated again by experts and eventually data set was proposed. Results: In first phase, based on the literature review, a draft set of 85 data elements was designed. In the second phase this data set was evaluated by experts and supplementary information was offered by professionals in subgroups especially in treatment part. In this phase the number of elements totally were arrived to 93 numbers. In the third phase, evaluation was conducted by experts and finally this dataset was designed in five main parts including: demographic information, diagnostic information, treatment information, clinical status assessment information, and clinical trial information. Conclusion: In this study the comprehensive core data set of colorectal cancer was designed. This dataset in the field of collecting colorectal cancer information can be useful through facilitating exchange of health information. Designing such data set for similar disease can help providers to collect standard data from patients and can accelerate retrieval from storage systems.

  8. Design Requirements of an Advanced HANARO Reactor Core Cooling System

    International Nuclear Information System (INIS)

    Park, Yong Chul; Ryu, Jeong Soo

    2007-12-01

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

  9. Neutronic design of the RSG-GAS silicide core

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, T.M.; Kuntoro, I.; Hastowo, H. [Center for Development of Research Reactor Technology National Nuclear Energy Agency BATAN, PUSPIPTEK Serpong Tangerang, 15310 (Indonesia)

    2002-07-01

    The objective of core conversion program of the RSG-GAS multipurpose reactor is to convert the fuel from oxide, U{sub 3}O{sub 8}-Al to silicide, U{sub 3}Si{sub 2}-Al. The aim of the program is to gain longer operation cycle by having, which is technically possible for silicide fuel, a higher density. Upon constraints of the existing reactor system and utilization, an optimal fuel density in amount of 3.55 g U/cc was found. This paper describes the neutronic parameter design of the silicide equilibrium core and the design of its transition cores as well. From reactivity control point of view, a modification of control rod system is also discussed. All calculations are carried out by means of diffusion codes, Batan-EQUIL-2D, Batan-2DIFF and -3DIFF. The silicide core shows that longer operation cycle of 32 full power days can be achieved without decreasing the safety criteria and utilization capabilities. (author)

  10. Conceptual Elements: A Detailed Framework to Support and Assess Student Learning of Biology Core Concepts

    Science.gov (United States)

    Cary, Tawnya; Branchaw, Janet

    2017-01-01

    The Vision and Change in Undergraduate Biology Education: Call to Action report has inspired and supported a nationwide movement to restructure undergraduate biology curricula to address overarching disciplinary concepts and competencies. The report outlines the concepts and competencies generally but does not provide a detailed framework to guide the development of the learning outcomes, instructional materials, and assessment instruments needed to create a reformed biology curriculum. In this essay, we present a detailed Vision and Change core concept framework that articulates key components that transcend subdisciplines and scales for each overarching biological concept, the Conceptual Elements (CE) Framework. The CE Framework was developed using a grassroots approach of iterative revision and incorporates feedback from more than 60 biologists and undergraduate biology educators from across the United States. The final validation step resulted in strong national consensus, with greater than 92% of responders agreeing that each core concept list was ready for use by the biological sciences community, as determined by scientific accuracy and completeness. In addition, we describe in detail how educators and departments can use the CE Framework to guide and document reformation of individual courses as well as entire curricula. PMID:28450444

  11. The understanding of core pharmacological concepts among health care students in their final semester.

    Science.gov (United States)

    Aronsson, Patrik; Booth, Shirley; Hägg, Staffan; Kjellgren, Karin; Zetterqvist, Ann; Tobin, Gunnar; Reis, Margareta

    2015-12-29

    The overall aim of the study was to explore health care students´ understanding of core concepts in pharmacology. An interview study was conducted among twelve students in their final semester of the medical program (n = 4), the nursing program (n = 4), and the specialist nursing program in primary health care (n = 4) from two Swedish universities. The participants were individually presented with two pharmacological clinically relevant written patient cases, which they were to analyze and propose a solution to. Participants were allowed to use the Swedish national drug formulary. Immediately thereafter the students were interviewed about their assessments. The interviews were audio-recorded and transcribed verbatim. A thematic analysis was used to identify units of meaning in each interview. The units were organized into three clusters: pharmacodynamics, pharmacokinetics, and drug interactions. Subsequent procedure consisted of scoring the quality of students´ understanding of core concepts. Non-parametric statistics were employed. The study participants were in general able to define pharmacological concepts, but showed less ability to discuss the meaning of the concepts in depth and to implement these in a clinical context. The participants found it easier to grasp concepts related to pharmacodynamics than pharmacokinetics and drug interactions. These results indicate that education aiming to prepare future health care professionals for understanding of more complex pharmacological reasoning and decision-making needs to be more focused and effective.

  12. Overview of the TITAN-II reversed-field pinch aqueous fusion power core design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Creedon, R.L.; Grotz, S.; Cheng, E.T.; Sharafat, S.; Cooke, P.I.H.

    1988-03-01

    TITAN-II is a compact, high power density Reversed-Field Pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO/sub 3/ and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the Reversed-Field Pinch reactor to operate at a neutron wall loading of 18 MWm/sup 2/. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a passive safety assurance design. 13 refs., 3 figs., 1 tab.

  13. Overview of the TITAN-II reversed-field pinch aqueous fusion power core design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Creedon, R.L.; Cheng, E.T. (General Atomic Co., San Diego, CA (USA)); Grotz, S.P.; Sharafat, S.; Cooke, P.I.H. (California Univ., Los Angeles (USA). Dept. of Mechanical, Aerospace and Nuclear Engineering; California Univ., Los Angeles, CA (USA). Inst. for Plasma and Fusion Research); TITAN Research Group

    1989-04-01

    TITAN-II is a compact, high-power-density Reversed-Field Pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO/sub 3/ and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the Reversed-Field Pinch reactor to operate at a neutron wall loading of 18 MW/m/sup 2/. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a level 2 of passive safety assurance design. (orig.).

  14. Design Concept Evaluation Using System Throughput Model

    International Nuclear Information System (INIS)

    Sequeira, G.; Nutt, W. M.

    2004-01-01

    The U.S. Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is currently developing the technical bases to support the submittal of a license application for construction of a geologic repository at Yucca Mountain, Nevada to the U.S. Nuclear Regulatory Commission. The Office of Repository Development (ORD) is responsible for developing the design of the proposed repository surface facilities for the handling of spent nuclear fuel and high level nuclear waste. Preliminary design activities are underway to sufficiently develop the repository surface facilities design for inclusion in the license application. The design continues to evolve to meet mission needs and to satisfy both regulatory and program requirements. A system engineering approach is being used in the design process since the proposed repository facilities are dynamically linked by a series of sub-systems and complex operations. In addition, the proposed repository facility is a major system element of the overall waste management process being developed by the OCRWM. Such an approach includes iterative probabilistic dynamic simulation as an integral part of the design evolution process. A dynamic simulation tool helps to determine if: (1) the mission and design requirements are complete, robust, and well integrated; (2) the design solutions under development meet the design requirements and mission goals; (3) opportunities exist where the system can be improved and/or optimized; and (4) proposed changes to the mission, and design requirements have a positive or negative impact on overall system performance and if design changes may be necessary to satisfy these changes. This paper will discuss the type of simulation employed to model the waste handling operations. It will then discuss the process being used to develop the Yucca Mountain surface facilities model. The latest simulation model and the results of the simulation and how the data were used in the design

  15. Pharmacology Goes Concept-Based: Course Design, Implementation, and Evaluation.

    Science.gov (United States)

    Lanz, Amelia; Davis, Rebecca G

    Although concept-based curricula are frequently discussed in the nursing education literature, little information exists to guide the development of a concept-based pharmacology course. Traditionally, nursing pharmacology courses are taught with an emphasis on drug class where a prototype drug serves as an exemplar. When transitioning pharmacology to a concept-based course, special considerations are in order. How can educators successfully integrate essential pharmacological content into a curriculum structured around nursing concepts? This article presents one approach to the design and implementation of a concept-based undergraduate pharmacology course. Planning methods, supportive teaching strategies, and course evaluation procedures are discussed.

  16. Engineering study for ISSTRS design concept

    Energy Technology Data Exchange (ETDEWEB)

    Hertzel, J.S.

    1997-01-31

    Los Alamos Technical Associates, Inc., is pleased to transmit the attached Conceptual Design Package for the Initial Single Shell Tank Retrieval System (ISSTRS), 90% Conceptual Design Review. The package includes the following: (1) ISSTRS Trade Studies: (a) Retrieval Facility Cooling Requirements; (b) Equipment Re-usability between Project W-320 and Tanks 241-C-103 and 241-C-1 05; (c) Sluice Line Options; and (d) Options for the Location of Tanks AX-103 and A-1 02 HVAC Equipment; (2) Drawings; (3) Risk Management Plan; (4) 0850 Interface Control Document; (5) Requirements Traceability Report; and (6) Project Design Specification.

  17. Concept and design of super junction devices

    Science.gov (United States)

    Zhang, Bo; Zhang, Wentong; Qiao, Ming; Zhan, Zhenya; Li, Zhaoji

    2018-02-01

    The super junction (SJ) has been recognized as the " milestone” of the power MOSFET, which is the most important innovation concept of the voltage-sustaining layer (VSL). The basic structure of the SJ is a typical junction-type VSL (J-VSL) with the periodic N and P regions. However, the conventional VSL is a typical resistance-type VSL (R-VSL) with only an N or P region. It is a qualitative change of the VSL from the R-VSL to the J-VSL, introducing the bulk depletion to increase the doping concentration and optimize the bulk electric field of the SJ. This paper firstly summarizes the development of the SJ, and then the optimization theory of the SJ is discussed for both the vertical and the lateral devices, including the non-full depletion mode, the minimum specific on-resistance optimization method and the equivalent substrate model. The SJ concept breaks the conventional " silicon limit” relationship of R on∝V B 2.5, showing a quasi-linear relationship of R on∝V B 1.03.

  18. Designing a Multichannel Map Service Concept

    Directory of Open Access Journals (Sweden)

    Hanna-Marika Halkosaari

    2013-01-01

    Full Text Available This paper introduces a user-centered design process for developing a multichannel map service. The aim of the service is to provide hikers with interactive maps through several channels. In a multichannel map service, the same spatial information is available through various channels, such as printed maps, Web maps, mobile maps, and other interactive media. When properly networked, the channels share a uniform identity so that the user experiences the different channels as a part of a single map service. The traditional methods of user-centered design, such as design probes, personas, and scenarios, proved useful even in the emerging field of developing multichannel map services. The findings emphasize the need to involve users and multidisciplinary teams in the conceptual phases of designing complex services aimed at serving various kinds of users.

  19. Core compressor exit stage study. 1: Aerodynamic and mechanical design

    Science.gov (United States)

    Burdsall, E. A.; Canal, E., Jr.; Lyons, K. A.

    1979-01-01

    The effect of aspect ratio on the performance of core compressor exit stages was demonstrated using two three stage, highly loaded, core compressors. Aspect ratio was identified as having a strong influence on compressors endwall loss. Both compressors simulated the last three stages of an advanced eight stage core compressor and were designed with the same 0.915 hub/tip ratio, 4.30 kg/sec (9.47 1bm/sec) inlet corrected flow, and 167 m/sec (547 ft/sec) corrected mean wheel speed. The first compressor had an aspect ratio of 0.81 and an overall pressure ratio of 1.357 at a design adiabatic efficiency of 88.3% with an average diffusion factor or 0.529. The aspect ratio of the second compressor was 1.22 with an overall pressure ratio of 1.324 at a design adiabatic efficiency of 88.7% with an average diffusion factor of 0.491.

  20. Ergonomic approach for pillow concept design.

    Science.gov (United States)

    Cai, Dengchuan; Chen, Hsiao-Lin

    2016-01-01

    Sleep quality is an essential factor to human beings for health. The current paper conducted four studies to provide a suitable pillow for promoting sleep quality. Study 1 investigated the natural positions of 40 subjects during sleep to derive key-points for a pillow design. The results suggested that the supine and lateral positions were alternatively 24 times a night, and the current pillows were too high for the supine position and too low for lateral positions. Study 2 measured body dimensions related to pillow design of 40 subjects to determine pillow sizes. The results suggested that the pillow height were quite different in supine position and lateral position and needed to take into consideration for a pillow design. Study 3 created a pillow design based on the results of above studies. The pillow was a U-form in the front of view in which the pillow height in the middle area was lower for the supine position, and both sides were higher for the lateral positions. Study 4 assessed sleep quality of 6 subjects by using the proposed pillows and the current pillows. The results showed that the newly designed pillow led to significantly higher sleep quality, and the new design received an innovation patent. Copyright © 2015 Elsevier Ltd and The Ergonomics Society. All rights reserved.

  1. The content and nature of a design concept

    DEFF Research Database (Denmark)

    Hansen, Claus Thorp; Andreasen, Mogens Myrup

    2002-01-01

    According to the design methodology literature "conceptual design" and "concepts" increase the effectiveness and efficiency of the early phases of a product development project, because conceptual thinking allows the engineering designer to identify or synthesise new unique solutions and allows him...... to focus his attention on the relatively few characteristics concerning the product´s functionality, and thereby makes it easier for the engineering designer to create several solution alternatives. In this paper we argue the following: 1. A conceptual design, i.e. the concept for a new product, may...... be seen from two sides, a need/market-oriented and a design/realisation-oriented. The need/market-oriented side explains the conceptual new way the design solves its task. The design/realisation side explains how the concept creates the necessary functionality and structural realisation for doing so. 2...

  2. Use of the Human Centered Design concept when designing ergonomic NPP control rooms

    International Nuclear Information System (INIS)

    Skrehot, Petr A.; Houser, Frantisek; Riha, Radek; Tuma, Zdenek

    2015-01-01

    Human-Centered Design is a concept aimed at reconciling human needs on the one hand and limitations posed by the design disposition of the room being designed on the other hand. This paper describes the main aspects of application of the Human-Centered Design concept to the design of nuclear power plant control rooms. (orig.)

  3. Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

    International Nuclear Information System (INIS)

    Philbin, J.S.; Nelson, W.E.; Rosenstroch, B.

    1978-10-01

    The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO 2 fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject

  4. Hyper-heuristic applied to nuclear reactor core design

    International Nuclear Information System (INIS)

    Domingos, R P; Platt, G M

    2013-01-01

    The design of nuclear reactors gives rises to a series of optimization problems because of the need for high efficiency, availability and maintenance of security levels. Gradient-based techniques and linear programming have been applied, as well as genetic algorithms and particle swarm optimization. The nonlinearity, multimodality and lack of knowledge about the problem domain makes de choice of suitable meta-heuristic models particularly challenging. In this work we solve the optimization problem of a nuclear reactor core design through the application of an optimal sequence of meta-heuritics created automatically. This combinatorial optimization model is known as hyper-heuristic.

  5. Simulated annealing algorithm for reactor in-core design optimizations

    International Nuclear Information System (INIS)

    Zhong Wenfa; Zhou Quan; Zhong Zhaopeng

    2001-01-01

    A nuclear reactor must be optimized for in core fuel management to make full use of the fuel, to reduce the operation cost and to flatten the power distribution reasonably. The author presents a simulated annealing algorithm. The optimized objective function and the punishment function were provided for optimizing the reactor physics design. The punishment function was used to practice the simulated annealing algorithm. The practical design of the NHR-200 was calculated. The results show that the K eff can be increased by 2.5% and the power distribution can be flattened

  6. Sound-proof Sandwich Panel Design via Metamaterial Concept

    Science.gov (United States)

    Sui, Ni

    Sandwich panels consisting of hollow core cells and two face-sheets bonded on both sides have been widely used as lightweight and strong structures in practical engineering applications, but with poor acoustic performance especially at low frequency regime. Basic sound-proof methods for the sandwich panel design are spontaneously categorized as sound insulation and sound absorption. Motivated by metamaterial concept, this dissertation presents two sandwich panel designs without sacrificing weight or size penalty: A lightweight yet sound-proof honeycomb acoustic metamateiral can be used as core material for honeycomb sandwich panels to block sound and break the mass law to realize minimum sound transmission; the other sandwich panel design is based on coupled Helmholtz resonators and can achieve perfect sound absorption without sound reflection. Based on the honeycomb sandwich panel, the mechanical properties of the honeycomb core structure were studied first. By incorporating a thin membrane on top of each honeycomb core, the traditional honeycomb core turns into honeycomb acoustic metamaterial. The basic theory for such kind of membrane-type acoustic metamaterial is demonstrated by a lumped model with infinite periodic oscillator system, and the negative dynamic effective mass density for clamped membrane is analyzed under the membrane resonance condition. Evanescent wave mode caused by negative dynamic effective mass density and impedance methods are utilized to interpret the physical phenomenon of honeycomb acoustic metamaterials at resonance. The honeycomb metamaterials can extraordinarily improve low-frequency sound transmission loss below the first resonant frequency of the membrane. The property of the membrane, the tension of the membrane and the numbers of attached membranes can impact the sound transmission loss, which are observed by numerical simulations and validated by experiments. The sandwich panel which incorporates the honeycomb metamateiral as

  7. Nuclear design and analysis report for KALIMER breakeven core conceptual design

    International Nuclear Information System (INIS)

    Kim, Sang Ji; Song, Hoon; Lee, Ki Bog; Chang, Jin Wook; Hong, Ser Gi; Kim, Young Gyun; Kim, Yeong Il

    2002-04-01

    During the phase 2 of LMR design technology development project, the breakeven core configuration was developed with the aim of the KALIMER self-sustaining with regard to the fissile material. The excess fissile material production is limited only to the extent of its own requirement for sustaining its planned power operation. The average breeding ratio is estimated to be 1.05 for the equilibrium core and the fissile plutonium gain per cycle is 13.9 kg. The nuclear performance characteristics as well as the reactivity coefficients have been analyzed so that the design evaluation in other activity areas can be made. In order to find out a realistic heavy metal flow evolution and investigate cycle-dependent nuclear performance parameter behaviors, the startup and transition cycle loading strategies are developed, followed by the startup core physics analysis. Driver fuel and blankets are assumed to be shuffled at the time of each reload. The startup core physics analysis has shown that the burnup reactivity swing, effective delayed neutron fraction, conversion ratio and peak linear heat generation rate at the startup core lead to an extreme of bounding physics data for safety analysis. As an outcome of this study, a whole spectrum of reactor life is first analyzed in detail for the KALIMER core. It is experienced that the startup core analysis deserves more attention than the current design practice, before the core configuration is finalized based on the equilibrium cycle analysis alone.

  8. Preliminary design concepts of an advanced integral reactor

    International Nuclear Information System (INIS)

    Moon, Kap S.; Lee, Doo J.; Kim, Keung K.; Chang, Moon H.; Kim, Si H.

    1997-01-01

    An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

  9. Concepts for passive heat removal and filtration systems under core meltdown conditions

    International Nuclear Information System (INIS)

    Wilhelm, J.G.; Neitzel, H.-J.

    1993-01-01

    The objective of the new containment concept being developed by KfK is the complete passive enclosure of a power reactor after a core meltdown accident by means of a solid containment structure and passive removal of the decay heat. This is to be accomplished by cooling the containment walls with ambient air, with thermoconvection as the driving force. The concept of the containment is described. Data are given of the heat removal and the requirements for filtration of the exhaust air, which is contaminated due to the leak rate assumed for the inner containment. The concept for the filter system is described. Various solutions for reduction of the large volumetric flow to be filtered are discussed. 3 refs., 8 figs

  10. A theoretical concept for a thermal-hydraulic 3D parallel channel core model

    International Nuclear Information System (INIS)

    Hoeld, A.

    2004-01-01

    A detailed description of the theoretical concept of the 3D thermal-hydraulic single- and two-phase flow phenomena is presented. The theoretical concept is based on important development lines such as separate treatment of the mass and energy from the momentum balance eqs. The other line is the establishment of a procedure for the calculation of the mass flow distributions into different parallel channels based on the fact that the sum of pressure decrease terms over a closed loop must stay, despite of un-symmetric perturbations, zero. The concept is realized in the experimental code HERO-X3D, concentrating in a first step on an artificial BWR or PWR core which may consist of a central channel, four quadrants, and a bypass channel. (authors)

  11. Remote metrology system (RMS) design concept

    International Nuclear Information System (INIS)

    1995-01-01

    A 3D remote metrology system (RMS) is needed to map the interior plasma-facing components of the International Thermonuclear Experimental Reactor (ITER). The performance and survival of these components within the reactor vessel are strongly dependent on their precise alignment and positioning with respect to the plasma edge. Without proper positioning and alignment, plasma-facing surfaces will erode rapidly. A RMS design involving Coleman Research Corporation (CRC) fiber optic coherent laser radar (CLR) technology is examined in this study. The fiber optic CLR approach was selected because its high precision should be able to meet the ITER 0.1 mm accuracy requirement and because the CLR's fiber optic implementation allows a 3D scanner to operate remotely from the RMS system's vulnerable components. This design study has largely verified that a fiber optic CLR based RMS can survive the ITER environment and map the ITER interior at the required accuracy at a one measurement/cm 2 density with a total measurement time of less than one hour from each of six or more vertically deployed measurement probes. The design approach employs a sealed and pressurized measurement probe which is attached with an umbilical spiral bellows conduit. This conduit bears fiber optic and electronic links plus a stream of air to lower the temperature in the interior of the probe. Lowering the probe temperature is desirable because probe electromechanical components which could survive the radiation environment often were not rated for the 200 C temperature. The tip of the probe whose outer shell has a flexible bellows joint can swivel in two degrees of freedom to allow mapping operations at each probe deployment level. This design study has concluded that the most successful scanner design will involve a hybrid AO beam deflector and mechanical scanner

  12. Remote metrology system (RMS) design concept

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-19

    A 3D remote metrology system (RMS) is needed to map the interior plasma-facing components of the International Thermonuclear Experimental Reactor (ITER). The performance and survival of these components within the reactor vessel are strongly dependent on their precise alignment and positioning with respect to the plasma edge. Without proper positioning and alignment, plasma-facing surfaces will erode rapidly. A RMS design involving Coleman Research Corporation (CRC) fiber optic coherent laser radar (CLR) technology is examined in this study. The fiber optic CLR approach was selected because its high precision should be able to meet the ITER 0.1 mm accuracy requirement and because the CLR`s fiber optic implementation allows a 3D scanner to operate remotely from the RMS system`s vulnerable components. This design study has largely verified that a fiber optic CLR based RMS can survive the ITER environment and map the ITER interior at the required accuracy at a one measurement/cm{sup 2} density with a total measurement time of less than one hour from each of six or more vertically deployed measurement probes. The design approach employs a sealed and pressurized measurement probe which is attached with an umbilical spiral bellows conduit. This conduit bears fiber optic and electronic links plus a stream of air to lower the temperature in the interior of the probe. Lowering the probe temperature is desirable because probe electromechanical components which could survive the radiation environment often were not rated for the 200 C temperature. The tip of the probe whose outer shell has a flexible bellows joint can swivel in two degrees of freedom to allow mapping operations at each probe deployment level. This design study has concluded that the most successful scanner design will involve a hybrid AO beam deflector and mechanical scanner.

  13. 76 FR 21789 - ITS Joint Program Office; Vehicle to Infrastructure Core System Concept of Operations; Notice of...

    Science.gov (United States)

    2011-04-18

    ... Concept of Operations; Notice of Public Meeting AGENCY: Research and Innovative Technology Administration...) Core System Concept of Operations on May 17, 2011 at the Detroit Metro Airport Marriott, 30559 Flynn... to work originally performed under the Vehicle Infrastructure Integration Proof of Concept (VII POC...

  14. Water chemistry of Atucha II PHWVR. Design concepts and evolution

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Rodriguez, Ivanna; Duca, Jorge; Fernandez, Ricardo; Rico, Jorge

    2007-01-01

    Full text: Atucha II is a pressurized heavy water vessel reactor designed by Siemens-KWU, currently part of AREVA NP, of 745 MWe and similar to Atucha I, which has been in operation over 25 years. The primary heat transport system (PHTS) is composed by vertical channels (277-313 C degrees) that allocate the fuel elements while the moderator circuit is composed by a partially separated circuit (142-173 C degrees). The moderation power is transferred to the feedwater through the moderator heat exchangers (HX). These HXs operate as the last, high pressure water-steam cycle heaters as well. Materials (with exception of fuel channels and fuel sheaths which are made of zirconium alloys) are all austenitic steels while cobalt containing alloys have been all replaced at the design stage. Steam generator and moderator HX tubing are Alloy 800 made. The core is operated without boron except with the first fresh nucleus. The secondary circuit or Balance of plant (BOP) is similar in conception to that of a PWR but the moderator HXs. It is entirely built of ferrous alloys, has a feedwater-deaerator tank and moisture separator. The energy sink is the Rio de la Plata River. The Reactors Chemistry Department, Chemistry Division, National Atomic Energy Commission, in its character of R and D institution has been committed by CNA II-N.A.S.A Project to prepare the water chemistry specifications, water chemistry engineering and manuals, considering the type of reactor, design and construction aspects and operation characteristics, taking into account the current state-of-the art and worldwide standards. This includes conceptual aspects and implementation and operative aspects as well. This documentation will be released after a designer's review as it has been stated in the respective agreement. Respecting the confidentiality agreement between CNEA and NASA and the confidentiality regarding handling original documentation provided by the designer, it is considered illustrative to

  15. MOX - equilibrium core design and trial irradiation in KAPS - 1

    International Nuclear Information System (INIS)

    Pradhan, A.S.; Ray, Sherly; Kumar, A.N.; Parikh, M.V.

    2006-01-01

    Option of usage of MOX fuel bundles in the equilibrium core of Indian 220 MWe PHWRs on a regular basis has been studied. The design of the MOX bundle considered is MOX -7 with inner 7 elements with uranium and plutonium oxide MOX fuel and outer 12 elements with natural uranium fuel. The composition of the plutonium isotopes corresponds to that at about 6500 MWD/TeU burnup. Burnup optimization has been done such that operation at design rated power is possible while achieving the maximum average discharge burnup. Operation with the optimized burnup pattern will result in substantial saving of natural uranium bundles. To obtain feedback on the performance of MOX bundles prior to its large scale use about 50 MOX-7 bundles have been loaded in KAPS - 1 equilibrium core. Locations have been selected such that reactor should be operating at rated power without violating any constraints on channel bundle powers and also meeting the safety requirements. Burnup of interest also should be achieved in minimum period of time. The fissile plutonium content in the 50 MOX fuel bundles loaded is about 75.6 wt % . About 38 bundles out of the 50 bundles loaded have been already discharged and remaining bundles are still in the core. The maximum discharge burnup of the MOX bundles is about 12000 MWD/TeU. The performance of the MOX bundles were excellent and as per prediction. No MOX bundle is reported to be failed. (author)

  16. Preliminary design of the new Proton Synchrotron Internal Dump core

    CERN Document Server

    AUTHOR|(CDS)2091975; Nuiry, François-Xavier

    The luminosity of the LHC particle accelerator at CERN is planned to be upgraded in the first half of 2020s, requiring also the upgrade of its injector accelerators, including the Proton Synchrotron (PS). The PS Internal Dumps are beam dumps located in the PS accelerator ring. They are safety devices designed to stop the circulating proton beam in order to protect the accelerator from damage due to an uncontrolled beam loss. The PS Internal Dumps need to be upgraded to be able to withstand the future higher intensity and energy proton beams. The dump core is a block of material interacting with the beam. It is located in ultra-high vacuum and moved into the beam path in 150 milliseconds by an electromagnet and spring-based actuation mechanism. The circulating proton beam is shaved by the core surface during thousands of beam revolutions. The preliminary new dump core design weighs 13 kilograms and consists of an isostatically pressed fine-grain graphite and a precipitation hardened copper alloy CuCrZr. The ...

  17. A concept for global optimization of topology design problems

    DEFF Research Database (Denmark)

    Stolpe, Mathias; Achtziger, Wolfgang; Kawamoto, Atsushi

    2006-01-01

    We present a concept for solving topology design problems to proven global optimality. We propose that the problems are modeled using the approach of simultaneous analysis and design with discrete design variables and solved with convergent branch and bound type methods. This concept is illustrated...... on two applications. The first application is the design of stiff truss structures where the bar areas are chosen from a finite set of available areas. The second considered application is simultaneous topology and geometry design of planar articulated mechanisms. For each application we outline...

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

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  19. Proximity operations concept design study, task 6

    Science.gov (United States)

    Williams, A. N.

    1990-01-01

    The feasibility of using optical technology to perform the mission of the proximity operations communications subsystem on Space Station Freedom was determined. Proximity operations mission requirements are determined and the relationship to the overall operational environment of the space station is defined. From this information, the design requirements of the communication subsystem are derived. Based on these requirements, a preliminary design is developed and the feasibility of implementation determined. To support the Orbital Maneuvering Vehicle and National Space Transportation System, the optical system development is straightforward. The requirements on extra-vehicular activity are such as to allow large fields of uncertainty, thus exacerbating the acquisition problem; however, an approach is given that could mitigate this problem. In general, it is found that such a system could indeed perform the proximity operations mission requirement, with some development required to support extra-vehicular activity.

  20. Review: BNL Tokamak graphite blanket design concepts

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1976-01-01

    The BNL minimum activity graphite blanket designs are reviewed, and three are discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a 30 cm or thicker graphite screen. Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy, which is then radiated to a secondary blanket with coolant tubes, as in types A and B, or removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma. (Auth.)

  1. UTN's gamma irradiation facility: design and concept

    International Nuclear Information System (INIS)

    Mohamad Noor Mohamad Yunus

    1986-01-01

    UTN is building a multipurpose gamma irradiation facility which compromises of research and pilot scale irradiation cells in The Fifth Malaysia Plan. The paper high-lights the basic futures of the facility in terms of its design and selection including layout sketches. Plant performances and limitations are discussed. Plants safety is briefly highlighted in block diagrams. Lastly, a typical specification brief is tabled in appendix for reference purposes. (author)

  2. New concept of microprocessor protective devices design

    Directory of Open Access Journals (Sweden)

    Gurevich Vladimir

    2010-01-01

    Full Text Available Transition from electromechanical to digital protective relays is accompanied with serious technical problems. The author offers a new approach in designing the digital relays capable of solving these problems. It is proposed to construct digital relays in the form of standard modules from which it would be possible to assemble the digital relay in the same way as now a personal computer.

  3. High field dipole magnet design concepts

    International Nuclear Information System (INIS)

    Nicol, T.H.

    1988-12-01

    High field dipole magnets will play a crucial role in the development of future accelerators whether at Fermilab or elsewhere. This paper presents conceptual designs for two such dipoles; 6.6 and 8.8 Tesla, with special focus on their suitability for upgrades to the Fermilab Tevatron. Descriptions and cross-sectional views will be presented as will preliminary estimates of heat loads and costs. 3 refs., 2 figs., 2 tabs

  4. Analysis and Multipoint Design of the TCA Concept

    Science.gov (United States)

    Krist, Steven E.; Bauer, Steven X. S.; Buning, Pieter G.

    1999-01-01

    The goal in this effort is to analyze the baseline TCA concept at transonic and supersonic cruise, then apply the natural flow wing design concept to obtain multipoint performance improvements. Analyses are conducted with OVERFLOW, a Navier-Stokes code for overset grids, using PEGSUS to compute the interpolations between the overset grids.

  5. Additive manufacturing for freeform mechatronics design: from concepts to applications

    NARCIS (Netherlands)

    Baars, G. van; Smeltink, J.; Werff, J. van der; Limpens, M.; Barink, M.; Berg, D. van den; Vreugd, J. de; Witvoet, G.; Galaktionov, O.S.

    2015-01-01

    This article presents developments of freeform mechatronics concepts, enabled by industrial Additive Manufacturing (AM), aiming at breakthroughs for precision engineering challenges such as lightweight, advanced thermal control, and integrated design. To assess potential impact in future

  6. Operational resilience: concepts, design and analysis

    Science.gov (United States)

    Ganin, Alexander A.; Massaro, Emanuele; Gutfraind, Alexander; Steen, Nicolas; Keisler, Jeffrey M.; Kott, Alexander; Mangoubi, Rami; Linkov, Igor

    2016-01-01

    Building resilience into today’s complex infrastructures is critical to the daily functioning of society and its ability to withstand and recover from natural disasters, epidemics, and cyber-threats. This study proposes quantitative measures that capture and implement the definition of engineering resilience advanced by the National Academy of Sciences. The approach is applicable across physical, information, and social domains. It evaluates the critical functionality, defined as a performance function of time set by the stakeholders. Critical functionality is a source of valuable information, such as the integrated system resilience over a time interval, and its robustness. The paper demonstrates the formulation on two classes of models: 1) multi-level directed acyclic graphs, and 2) interdependent coupled networks. For both models synthetic case studies are used to explore trends. For the first class, the approach is also applied to the Linux operating system. Results indicate that desired resilience and robustness levels are achievable by trading off different design parameters, such as redundancy, node recovery time, and backup supply available. The nonlinear relationship between network parameters and resilience levels confirms the utility of the proposed approach, which is of benefit to analysts and designers of complex systems and networks.

  7. Structural Design and Sizing of a Metallic Cryotank Concept

    Science.gov (United States)

    Sleight, David W.; Martin, Robert A.; Johnson, Theodore F.

    2013-01-01

    This paper presents the structural design and sizing details of a 33-foot (10 m) metallic cryotank concept used as the reference design to compare with the composite cryotank concepts developed by industry as part of NASA s Composite Cryotank Technology Development (CCTD) Project. The structural design methodology and analysis results for the metallic cryotank concept are reported in the paper. The paper describes the details of the metallic cryotank sizing assumptions for the baseline and reference tank designs. In particular, the paper discusses the details of the cryotank weld land design and analyses performed to obtain a reduced weight metallic cryotank design using current materials and manufacturing techniques. The paper also discusses advanced manufacturing techniques to spin-form the cryotank domes and compares the potential mass savings to current friction stir-welded technology.

  8. Development of MMIS design concepts for the KNGR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Ku, In Su; Heo, Seop; Jeong, Chel Hwan; Lee, Hyun Chol; Park, Hui Yun; Lee, Chol Gwon; So, Yong Suk; Kim, Dong Hun; Jang, Gwi Sook; Lee, Ki Yonug; Lee, Jun; Kim, Young In [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-12-01

    The design goals of MMIS for the next generation nuclear power plant are to improve plant safety and the cost effectiveness of nuclear power plants, and to meet with regulatory requirements. For the optimized design of MMIS, conceptual design bases are required for the optimization of MMIS design to establish the design concepts for NGR MMIS. The conceptual design bases are also required for performing the basic design, and verifying the design. The objectives of this study are establishment of MMIS design bases and the development of next generation MMIS configuration concepts. The MMIS design bases for by adopting MMIS requirements developed in the previous study on next generation reactor evaluation techniques and advanced MMIS technologies. The next generation MMIS design requirements are to be developed based on the device obsolescence problems by applying modern digital technology. This report describes the design concepts for the next generation MMIS. In order to develop the design concepts, new technologies were analyzed, and the characteristics of new advanced MMIS designs were reviewed. In addition, reviewing the advanced design features (ADF) resulted from the 3 rd stage of standardization project, the strategy for the application of the results from these activities are prepared. This report includes the comparison results of the design characteristics of next generation MMIS with those of existing plants, YGN 3 and 4, UCN 3 and 4, and NUPLEX 80+. This report also describes the conceptual MMIS configuration of next generation control room, based on the results from the comparison. The results of this study will be an input for the detailed design guidelines and a regulatory requirements review report for the next generation MMIS design, and provide basis for the basis and detailed design of MMI and I and C for main control room. 1 fig., 1 tab., 46 refs. (Author) .new.

  9. Development of MMIS design concepts for the KNGR

    International Nuclear Information System (INIS)

    Park, Jong Kyun; Ku, In Su; Heo, Seop; Jeong, Chel Hwan; Lee, Hyun Chol; Park, Hui Yun; Lee, Chol Gwon; So, Yong Suk; Kim, Dong Hun; Jang, Gwi Sook; Lee, Ki Yonug; Lee, Jun; Kim, Young In

    1995-12-01

    The design goals of MMIS for the next generation nuclear power plant are to improve plant safety and the cost effectiveness of nuclear power plants, and to meet with regulatory requirements. For the optimized design of MMIS, conceptual design bases are required for the optimization of MMIS design to establish the design concepts for NGR MMIS. The conceptual design bases are also required for performing the basic design, and verifying the design. The objectives of this study are establishment of MMIS design bases and the development of next generation MMIS configuration concepts. The MMIS design bases for by adopting MMIS requirements developed in the previous study on next generation reactor evaluation techniques and advanced MMIS technologies. The next generation MMIS design requirements are to be developed based on the device obsolescence problems by applying modern digital technology. This report describes the design concepts for the next generation MMIS. In order to develop the design concepts, new technologies were analyzed, and the characteristics of new advanced MMIS designs were reviewed. In addition, reviewing the advanced design features (ADF) resulted from the 3 rd stage of standardization project, the strategy for the application of the results from these activities are prepared. This report includes the comparison results of the design characteristics of next generation MMIS with those of existing plants, YGN 3 and 4, UCN 3 and 4, and NUPLEX 80+. This report also describes the conceptual MMIS configuration of next generation control room, based on the results from the comparison. The results of this study will be an input for the detailed design guidelines and a regulatory requirements review report for the next generation MMIS design, and provide basis for the basis and detailed design of MMI and I and C for main control room. 1 fig., 1 tab., 46 refs. (Author) .new

  10. Feasibility study on thermal-hydraulic design of reduced-moderation PWR-type core

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

    2000-03-01

    At JAERI, a conceptual study on reduced-moderation water reactor (RMWR) has been performed as one of the advanced reactor system which is designed so as to realize the conversion ratio more than unity. In this reactor concept, the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated. Therefore, an evaluation of the core thermal margin becomes very important in the design of the RMWR. In this study, we have performed a feasibility evaluation on thermal-hydraulic design of RM-PWR type core (core thermal output: 2900 MWt, Rod gaps: 1 mm). In RM-PWR core, seed and blanket regions are exist. In the blanket region, power density is lower than that of the seed region. Then, evaluation was performed under setting a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because it is possible that the coolant boils in the seed region. In the feasibility evaluations, subchannel code COBRA-IV-I was used in combination with KfK DNB (departure nucleate boiling) correlation. When coolant mass flow rate to the blanket fuel assembly is reduced by 40%, and that to the seed fuel assembly is increased, coolant boiling is not occurred in the assembly region calculation. Provided that the channel boxes to the blanket fuel assembly are set up and coolant mass flow rate to the blanket fuel assembly is reduced by 40%, it is confirmed by the whole core calculation that the boiling of the coolant is not occurred and the RM-PWR core is feasible. (author)

  11. Fractal design concepts for stretchable electronics.

    Science.gov (United States)

    Fan, Jonathan A; Yeo, Woon-Hong; Su, Yewang; Hattori, Yoshiaki; Lee, Woosik; Jung, Sung-Young; Zhang, Yihui; Liu, Zhuangjian; Cheng, Huanyu; Falgout, Leo; Bajema, Mike; Coleman, Todd; Gregoire, Dan; Larsen, Ryan J; Huang, Yonggang; Rogers, John A

    2014-01-01

    Stretchable electronics provide a foundation for applications that exceed the scope of conventional wafer and circuit board technologies due to their unique capacity to integrate with soft materials and curvilinear surfaces. The range of possibilities is predicated on the development of device architectures that simultaneously offer advanced electronic function and compliant mechanics. Here we report that thin films of hard electronic materials patterned in deterministic fractal motifs and bonded to elastomers enable unusual mechanics with important implications in stretchable device design. In particular, we demonstrate the utility of Peano, Greek cross, Vicsek and other fractal constructs to yield space-filling structures of electronic materials, including monocrystalline silicon, for electrophysiological sensors, precision monitors and actuators, and radio frequency antennas. These devices support conformal mounting on the skin and have unique properties such as invisibility under magnetic resonance imaging. The results suggest that fractal-based layouts represent important strategies for hard-soft materials integration.

  12. Fractal design concepts for stretchable electronics

    Science.gov (United States)

    Fan, Jonathan A.; Yeo, Woon-Hong; Su, Yewang; Hattori, Yoshiaki; Lee, Woosik; Jung, Sung-Young; Zhang, Yihui; Liu, Zhuangjian; Cheng, Huanyu; Falgout, Leo; Bajema, Mike; Coleman, Todd; Gregoire, Dan; Larsen, Ryan J.; Huang, Yonggang; Rogers, John A.

    2014-02-01

    Stretchable electronics provide a foundation for applications that exceed the scope of conventional wafer and circuit board technologies due to their unique capacity to integrate with soft materials and curvilinear surfaces. The range of possibilities is predicated on the development of device architectures that simultaneously offer advanced electronic function and compliant mechanics. Here we report that thin films of hard electronic materials patterned in deterministic fractal motifs and bonded to elastomers enable unusual mechanics with important implications in stretchable device design. In particular, we demonstrate the utility of Peano, Greek cross, Vicsek and other fractal constructs to yield space-filling structures of electronic materials, including monocrystalline silicon, for electrophysiological sensors, precision monitors and actuators, and radio frequency antennas. These devices support conformal mounting on the skin and have unique properties such as invisibility under magnetic resonance imaging. The results suggest that fractal-based layouts represent important strategies for hard-soft materials integration.

  13. Developing engineering design core competences through analysis of industrial products

    DEFF Research Database (Denmark)

    Hansen, Claus Thorp; Lenau, Torben Anker

    2011-01-01

    Most product development work carried out in industrial practice is characterised by being incremental, i.e. the industrial company has had a product in production and on the market for some time, and now time has come to design a new and upgraded variant. This type of redesign project requires...... that the engineering designers have core design competences to carry through an analysis of the existing product encompassing both a user-oriented side and a technical side, as well as to synthesise solution proposals for the new and upgraded product. The authors of this paper see an educational challenge in staging...... a course module, in which students develop knowledge, understanding and skills, which will prepare them for being able to participate in and contribute to redesign projects in industrial practice. In the course module Product Analysis and Redesign that has run for 8 years we have developed and refined...

  14. Trajectory Design for a Single-String Impactor Concept

    Science.gov (United States)

    Dono Perez, Andres; Burton, Roland; Stupl, Jan; Mauro, David

    2017-01-01

    This paper introduces a trajectory design for a secondary spacecraft concept to augment science return in interplanetary missions. The concept consist of a single-string probe with a kinetic impactor on board that generates an artificial plume to perform in-situ sampling. The trajectory design was applied to a particular case study that samples ejecta particles from the Jovian moon Europa. Results were validated using statistical analysis. Details regarding the navigation, targeting and disposal challenges related to this concept are presented herein.

  15. BWR power oscillation evaluation methodologies in core design

    International Nuclear Information System (INIS)

    Hotta, Akitoshi

    1995-01-01

    At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

  16. LFR core design for prevention & mitigation of severe accidents

    International Nuclear Information System (INIS)

    Grasso, Giacomo

    2012-01-01

    Conclusions: • Aiming at fully complying Gen-IV safety requirements – even in case of Fukushima-like events –, prevention and mitigation strategies must be stressed in FR design. • The safety of Lead-cooled Fast Reactors can rely on intrinsic features due to the coolant, such as: • the practical impossibility of Lead boiling, hence the unreliability of core (only) voiding for wide safety margins, and the retention of corium; • the high density of lead, for the buoyancy of Control Rods (allowing their safe positioning below the core), and the dispersion of molten core up to the setting up of a “cold melting pot”. • the possibility to adopt wide coolant channels for encouraging natural circulation, without affecting the hardness of the neutron spectrum; • the hard neutron spectrum allows the adiabatic operation of LFRs (which implies minimal criticality swings even through long cycles) with small amounts of Mas (hence with a negligible detriment to the safety features); • an effective reduction of the coolant density effect simply through the shortening of the active height

  17. The SSC superconducting air core toroid design development

    International Nuclear Information System (INIS)

    Fields, T.; Carroll, A.; Chiang, I.H.; Frank, J.S.; Haggerty, J.; Littenberg, L.; Morse, W.; Strand, R.C.; Lau, K.; Weinstein, R.; McNeil, R.; Friedman, J.; Hafen, E.; Haridas, P.; Kendall, H.W.; Osborne, L.; Pless, I.; Rosenson, L.; Pope, B.; Jones, L.W.; Luton, J.N.; Bonanos, P.; Marx, M.; Pusateri, J.A.; Favale, A.; Gottesman, S.; Schneid, E.; Verdier, R.

    1990-01-01

    Superconducting air core toroids show great promise for use in a muon spectrometer for the SSC. Early studies by SUNY at Stony Brook funded by SSC Laboratory, have established the feasibility of building magnets of the required size. The toroid spectrometer consists of a central toroid with two end cap toroids. The configuration under development provides for muon trajectory measurement outside the magnetic volume. System level studies on support structure, assembly, cryogenic material selection, and power are performed. Resulting selected optimal design and assembly is described. 4 refs., 6 figs

  18. Space Launch System, Core Stage, Structural Test Design and Implementation

    Science.gov (United States)

    Shaughnessy, Ray

    2017-01-01

    As part of the National Aeronautics and Space Administration's (NASA) Space Launch System (SLS) Program, engineers at NASA's Marshall Space Flight Center (MSFC) in Huntsville, Alabama are working to design, develop and implement the SLS Core Stage structural testing. The SLS will have the capability to return humans to the Moon and beyond and its first launch is scheduled for December of 2017. The SLS Core Stage consist of five major elements; Forward Skirt, Liquid Oxygen (LOX) tank, Intertank (IT), Liquid Hydrogen (LH2) tank and the Engine Section (ES). Structural Test Articles (STA) for each of these elements are being designed and produced by Boeing at Michoud Assembly Facility located in New Orleans, La. The structural test for the Core Stage STAs (LH2, LOX, IT and ES) are to be conducted by the MSFC Test Laboratory. Additionally, the MSFC Test Laboratory manages the Structural Test Equipment (STE) design and development to support the STAs. It was decided early (April 2012) in the project life that the LH2 and LOX tank STAs would require new test stands and the Engine Section and Intertank would be tested in existing facilities. This decision impacted schedules immediately because the new facilities would require Construction of Facilities (C of F) funds that require congressional approval and long lead times. The Engine Section and Intertank structural test are to be conducted in existing facilities which will limit lead times required to support the first launch of SLS. With a SLS launch date of December, 2017 Boeing had a need date for testing to be complete by September of 2017 to support flight certification requirements. The test facilities were required to be ready by October of 2016 to support test article delivery. The race was on to get the stands ready before Test Article delivery and meet the test complete date of September 2017. This paper documents the past and current design and development phases and the supporting processes, tools, and

  19. Core damage frequency (reactor design) perspectives based on IPE results

    International Nuclear Information System (INIS)

    Camp, A.L.; Dingman, S.E.; Forester, J.A.

    1996-01-01

    This paper provides perspectives gained from reviewing 75 Individual Plant Examination (IPE) submittals covering 108 nuclear power plant units. Variability both within and among reactor types is examined to provide perspectives regarding plant-specific design and operational features, and C, modeling assumptions that play a significant role in the estimates of core damage frequencies in the IPEs. Human actions found to be important in boiling water reactors (BWRs) and in pressurized water reactors (PWRs) are presented and the events most frequently found important are discussed

  20. Habitable Exoplanet Imager Optical Telescope Concept Design

    Science.gov (United States)

    Stahl, H Philip

    2017-01-01

    The Habitable Exoplanet Imaging Mission (HabEx) is one of four missions under study for the 2020 Astrophysics Decadal Survey. Its goal is to directly image and spectroscopically characterize planetary systems in the habitable zone of Sun-like stars. Additionally, HabEx will perform a broad range of general astrophysics science enabled by 100 to 2500 nm spectral range and 3 x 3 arc-minute FOV. Critical to achieving the HabEx science goals is a large, ultra-stable UV/Optical/Near-IR (UVOIR) telescope. The baseline HabEx telescope is a 4-meter off-axis unobscured three-mirror-anastigmatic, diffraction limited at 400 nm with wavefront stability on the order of a few 10s of picometers. This paper summarizes the opto-mechanical design of the HabEx baseline optical telescope assembly, including a discussion of how science requirements drive the telescope's specifications, and presents analysis that the baseline telescope structure meets its specified tolerances.

  1. Thermal design trades for SAFIR architecture concepts

    Science.gov (United States)

    Yorke, Harold W.; Paine, Christopher; Bradford, Matt; Dragovan, Mark; Nash, Al; Dooley, Jennifer; Lawrence, Charles

    2004-01-01

    SAFIR is a IO-meter, 4 K space telescope optimized for wavelengths between 20 microns and 1 mm. The combination of aperture diameter and telescope temperature will provide a raw sensitivity improvement of more than a factor of 1000 over presently-planned missions. The sensitivity will be comparable to that of the JWST and ALMA, but at the critical far-IR wavelengths where much of the universe's radiative energy has emerged since the origin of stars and galaxies. We examine several of the critical technologies for SAFIR which enable the large cold aperture, and present results of studies examining the telescope optics and the spacecraft thermal architecture. Both the method by which the aperture is filled, and the overall optical design for the telescope can impact the potential scientific return of SAFIR. Thermal architecture that goes far beyond the sunshades developed for the James Webb Space Telescope will be necessary to achieve the desired sensitivity of SAFIR. By combining active and passive cooling at critical points within the observatory, a significant reduction of the required level of active cooling can be obtained.

  2. Habitable exoplanet imager optical telescope concept design

    Science.gov (United States)

    Stahl, H. Philip

    2017-09-01

    The Habitable Exoplanet Imaging Mission (HabEx) is one of four missions under study for the 2020 Astrophysics Decadal Survey. Its goal is to directly image and spectroscopically characterize planetary systems in the habitable zone of Sunlike stars. Additionally, HabEx will perform a broad range of general astrophysics science enabled by 100 to 2500 nm spectral range and 3 x 3 arc-minute FOV. Critical to achieving the HabEx science goals is a large, ultra-stable UV/Optical/Near-IR (UVOIR) telescope. The baseline HabEx telescope is a 4-meter off-axis unobscured three-mirroranastigmatic, diffraction limited at 400 nm with wavefront stability on the order of a few 10s of picometers. This paper summarizes the opto-mechanical design of the HabEx baseline optical telescope assembly, including a discussion of how science requirements drive the telescope's specifications, and presents analysis that the baseline telescope structure meets its specified tolerances.

  3. Reactor design concepts for radiation processing

    International Nuclear Information System (INIS)

    Berejka, A.J.

    2004-01-01

    During the formative years of irradiation processing, the 1950s and 1960s, there was laboratory and academic interest in the use of this form of energy transfer to initiate polymerization for the manufacture of plastics and in other chemical processes. Studies were often based on low-dose-rate Cobalt-60 systems. The electron beam (EB) accelerator technology of the time was not as yet at the robust and industrially reliable state that it is now at the beginning of the twenty-first century. A series of reactor designs illustrate how an electron beam can be incorporated into reactor vessels for initiating gas and liquid phase polymerizations on a continuous basis. Development of such approaches, which would rely upon contemporary, high current electron beams to initiate polymerization, would help the chemical processing industry alleviate its problems of catalyst disposal and its related environmental concerns. Systems for treating materials in bulk at low doses, such as those typically used for grain disinfection, at high through-put rates, are also illustrated. Simplified shielding is envisioned in each proposed process system

  4. Applying and incorporating user driven innovation when designing concepts

    DEFF Research Database (Denmark)

    Thorp Hansen, Claus; Brønnum, Louise

    This paper addresses the difficulties seen when working within the user driven innovation [UDI] paradigm. We examine some of the circumstances that often make it difficult to work with user insights in concept design. UDI has become a recognized design approach, but has not yet accommodated...... a design practice explicitly considering the type of user insights this approach implies. For that reason UDI has yet to prove itself and its potential effect; a study of Danish initiative “program for user driven innovation” has shown little effect in this regard. However it has shown that radical new...... insights have been produced but at the same time to abstract when integrated in the design process. We will discuss and propose a framework for working with user insights in concept design, based on existing concept frameworks but actively addressing and incorporating user insights as a new type of input...

  5. Theoretical and numerical studies of TWR based on ESFR core design

    International Nuclear Information System (INIS)

    Zhang, Dalin; Chen, Xue-Nong; Flad, Michael; Rineiski, Andrei; Maschek, Werner

    2013-01-01

    Highlights: • The traveling wave reactor (TWR) is studied based on the core design of the European Sodium-cooled Fast Reactor (ESFR). • The conventional fuel shuffling technique is used to produce a continuous radial fuel movement. • A stationary self sustainable nuclear fission power can be established asymptotically by only loading natural or depleted uranium. • The multi-group deterministic neutronic code ERANOS is applied. - Abstract: This paper deals with the so-called traveling wave reactor (TWR) based on the core design of the European Sodium-cooled Fast Reactor (ESFR). The current concept of TWR is to use the conventional radial fuel shuffling technique to produce a continuous radial fuel movement so that a stationary self sustainable nuclear fission power can be established asymptotically by only loading fertile material consisting of natural or depleted uranium. The core design of ESFR loaded with metallic uranium fuel without considering the control mechanism is used as a practical application example. The theoretical studies focus mainly on qualitative feasibility analyses, i.e. to identify out in general essential parameter dependences of such a kind of reactor. The numerical studies are carried out more specifically on a certain core design. The multi-group deterministic neutronic code ERANOS with the JEFF3.1 data library is applied as a basic tool to perform the neutronics and burn-up calculations. The calculations are performed in a 2-D R-Z geometry, which is sufficient for the current core layout. Numerical results of radial fuel shuffling indicate that the asymptotic k eff parabolically varies with the shuffling period, while the burn-up increases linearly. Typical shuffling periods investigated in this study are in the range of 300–1000 days. The important parameters, e.g. k eff , the burn-up, the power peaking factor, and safety coefficients are calculated

  6. On-line generation of core monitoring power distribution in the SCOMS couppled with core design code

    International Nuclear Information System (INIS)

    Lee, K. B.; Kim, K. K.; In, W. K.; Ji, S. K.; Jang, M. H.

    2002-01-01

    The paper provides the description of the methodology and main program module of power distribution calculation of SCOMS(SMART COre Monitoring System). The simulation results of the SMART core using the developed SCOMS are included. The planar radial peaking factor(Fxy) is relatively high in SMART core because control banks are inserted to the core at normal operation. If the conventional core monitoring method is adapted to SMART, highly skewed planar radial peaking factor Fxy yields an excessive conservatism and reduces the operation margin. In addition to this, the error of the core monitoring would be enlarged and thus operating margin would be degraded, because it is impossible to precalculate the core monitoring constants for all the control banks configurations taking into account the operation history in the design stage. To get rid of these drawbacks in the conventional power distribution calculation methodology, new methodology to calculate the three dimensional power distribution is developed. Core monitoring constants are calculated with the core design code (MASTER) which is on-line coupled with SCOMS. Three dimensional (3D) power distribution and the several peaking factors are calculated using the in-core detector signals and core monitoring constant provided at real time. Developed methodology is applied to the SMART core and the various core states are simulated. Based on the simulation results, it is founded that the three dimensional peaking factor to calculate the Linear Power Density and the pseudo hot-pin axial power distribution to calculate the Departure Nucleate Boiling Ratio show the more conservative values than those of the best-estimated core design code, and SCOMS adapted developed methodology can secures the more operation margin than the conventional methodology

  7. A supercomputing application for reactors core design and optimization

    International Nuclear Information System (INIS)

    Hourcade, Edouard; Gaudier, Fabrice; Arnaud, Gilles; Funtowiez, David; Ammar, Karim

    2010-01-01

    Advanced nuclear reactor designs are often intuition-driven processes where designers first develop or use simplified simulation tools for each physical phenomenon involved. Through the project development, complexity in each discipline increases and implementation of chaining/coupling capabilities adapted to supercomputing optimization process are often postponed to a further step so that task gets increasingly challenging. In the context of renewal in reactor designs, project of first realization are often run in parallel with advanced design although very dependant on final options. As a consequence, the development of tools to globally assess/optimize reactor core features, with the on-going design methods accuracy, is needed. This should be possible within reasonable simulation time and without advanced computer skills needed at project management scale. Also, these tools should be ready to easily cope with modeling progresses in each discipline through project life-time. An early stage development of multi-physics package adapted to supercomputing is presented. The URANIE platform, developed at CEA and based on the Data Analysis Framework ROOT, is very well adapted to this approach. It allows diversified sampling techniques (SRS, LHS, qMC), fitting tools (neuronal networks...) and optimization techniques (genetic algorithm). Also data-base management and visualization are made very easy. In this paper, we'll present the various implementing steps of this core physics tool where neutronics, thermo-hydraulics, and fuel mechanics codes are run simultaneously. A relevant example of optimization of nuclear reactor safety characteristics will be presented. Also, flexibility of URANIE tool will be illustrated with the presentation of several approaches to improve Pareto front quality. (author)

  8. Overview of neutronic fuel assembly design and in-core fuel management

    International Nuclear Information System (INIS)

    Porsch, D.; Charlier, A.; Meier, G.; Mougniot, J.C.; Tsuda, K.

    2000-01-01

    The civil and military utilization of nuclear power results in stockpiles of spent fuel and separated plutonium. Recycling of the recovered plutonium in Light Water Reactors (LWR) is currently practiced in Belgium, France, Germany, and Switzerland, in Japan it is in preparation. Modern MOX fuel, with its optimized irradiation and reprocessing behavior, was introduced in 1981. Since then, about 1700 MOX fuel assemblies of different mechanical and neutronic design were irradiated in commercial LWRs and reached fuel assembly averaged exposures of up to 51.000 MWd/t HM. MOX fuel assemblies reloaded in PWR have an average fissile plutonium content of up to 4.8 w/o. For BWR, the average fissile plutonium content in actual reloads is 3.0 w/o. Targets for the MOX fuel assembly design are the compatibility to uranium fuel assemblies with respect to their mechanical fuel rod and fuel assembly design, they should have no impact on the flexibility of the reactor operation, and its reload should be economically feasible. In either cycle independent safety analyses or individually for each designed core it has to be demonstrated that recycling cores meet the same safety criteria as uranium cores. The safety criteria are determined for normal operation and for operational as well as design basis transients. Experience with realized MOX core loadings confirms the reliability of the applied modern design codes. Studies for reloads of advanced MOX assemblies in LWRs demonstrate the feasibility of a future development of the thermal plutonium recycling. New concepts for the utilization of plutonium are under consideration and reveal an attractive potential for further developments on the plutonium exploitation sector. (author)

  9. Modeling and simulation in the systems engineering life cycle core concepts and accompanying lectures

    CERN Document Server

    Loper, Margaret L

    2015-01-01

    This easy to read text/reference provides a broad introduction to the fundamental concepts of modeling and simulation (M&S) and systems engineering, highlighting how M&S is used across the entire systems engineering lifecycle. Each chapter corresponds to a short lecture covering a core topic in M&S or systems engineering.  Topics and features: reviews the full breadth of technologies, methodologies and uses of M&S, rather than just focusing on a specific aspect of the field; presents contributions from renowned specialists in each topic covered; introduces the foundational elements and proce

  10. Design and evaluation of a foam-filled hat-stiffened panel concept for aircraft primary structural applications

    Science.gov (United States)

    Ambur, Damodar R.

    1995-01-01

    A structurally efficient hat-stiffened panel concept that utilizes a structural foam as stiffener core has been designed for aircraft primary structural applications. This stiffener concept utilizes a manufacturing process that can be adapted readily to grid-stiffened structural configurations which possess inherent damage tolerance characteristics due to their multiplicity of load paths. The foam-filled hat-stiffener concept in a prismatically stiffened panel configuration is more efficient than most other stiffened panel configurations in a load range that is typical for both fuselage and wing structures. The prismatically stiffened panel concept investigated here has been designed using AS4/3502 preimpregnated tape and Rohacell foam core and evaluated for its buckling and postbuckling behavior with and without low-speed impact damage. The results from single-stiffener and multi-stiffener specimens suggest that this structural concept responds to loading as anticipated and has good damage tolerance characteristics.

  11. Design and Evaluation of Nextgen Aircraft Separation Assurance Concepts

    Science.gov (United States)

    Johnson, Walter; Ho, Nhut; Arutyunov, Vladimir; Laue, John-Luke; Wilmoth, Ian

    2012-01-01

    To support the development and evaluation of future function allocation concepts for separation assurance systems for the Next Generation Air Transportation System, this paper presents the design and human-in-the-loop evaluation of three feasible function allocation concepts that allocate primary aircraft separation assurance responsibilities and workload to: 1) pilots; 2) air traffic controllers (ATC); and 3) automation. The design of these concepts also included rules of the road, separation assurance burdens for aircraft of different equipage levels, and utilization of advanced weather displays paired with advanced conflict detection and resolution automation. Results of the human-in-the-loop simulation show that: a) all the concepts are robust with respect to weather perturbation; b) concept 1 (pilots) had highest throughput, closest to assigned spacing, and fewest violations of speed and altitude restrictions; c) the energy of the aircraft during the descent phase was better managed in concepts 1 and 2 (pilots and ATC) than in concept 3 (automation), in which the situation awareness of pilots and controllers was lowest, and workload of pilots was highest. The paper also discusses further development of these concepts and their augmentation and integration with future air traffic management tools and systems that are being considered for NextGen.

  12. Tools and applications for core design and shielding in fast reactors

    International Nuclear Information System (INIS)

    Rachamin, Reuven

    2013-01-01

    Outline: • Modeling of SFR cores using the Serpent-DYN3D code sequence; • Core shielding assessment for the design of FASTEF-MYRRHA; • Neutron shielding studies on an advanced Molten Salt Fast Reactor (MSFR) design

  13. A study on the advanced statistical core thermal design methodology

    International Nuclear Information System (INIS)

    Lee, Seung Hyuk

    1992-02-01

    A statistical core thermal design methodology for generating the limit DNBR and the nominal DNBR is proposed and used in assessing the best-estimate thermal margin in a reactor core. Firstly, the Latin Hypercube Sampling Method instead of the conventional Experimental Design Technique is utilized as an input sampling method for a regression analysis to evaluate its sampling efficiency. Secondly and as a main topic, the Modified Latin Hypercube Sampling and the Hypothesis Test Statistics method is proposed as a substitute for the current statistical core thermal design method. This new methodology adopts 'a Modified Latin Hypercube Sampling Method' which uses the mean values of each interval of input variables instead of random values to avoid the extreme cases that arise in the tail areas of some parameters. Next, the independence between the input variables is verified through 'Correlation Coefficient Test' for statistical treatment of their uncertainties. And the distribution type of DNBR response is determined though 'Goodness of Fit Test'. Finally, the limit DNBR with one-sided 95% probability and 95% confidence level, DNBR 95/95 ' is estimated. The advantage of this methodology over the conventional statistical method using Response Surface and Monte Carlo simulation technique lies in its simplicity of the analysis procedure, while maintaining the same level of confidence in the limit DNBR result. This methodology is applied to the two cases of DNBR margin calculation. The first case is the application to the determination of the limit DNBR where the DNBR margin is determined by the difference between the nominal DNBR and the limit DNBR. The second case is the application to the determination of the nominal DNBR where the DNBR margin is determined by the difference between the lower limit value of the nominal DNBR and the CHF correlation limit being used. From this study, it is deduced that the proposed methodology gives a good agreement in the DNBR results

  14. Control rod repositioning considerations in core design analysis

    International Nuclear Information System (INIS)

    Armstrong, B.C.; Buechel, R.J.

    1990-01-01

    Control rod repositioning is a method for minimizing rod cluster control assembly (RCCA) wear in the upper internals area where the guide cards interface with the rodlets of the RCCAs. A number of utilities have implemented strategies for rod repositioning, which often has no impact on the nuclear analysis for cases where the control rods are never repositioned into the active fuel. Other strategies involve repositioning the control rods several steps into the active fuel. The impact of this type of repositioning on the axial power shape and consequently the total peaking factor F Q T varies, depending on the method in which the repositioning is implemented at the plant. Operating for long periods with all the control and shutdown rods inserted several steps in the active fuel followed by withdrawing them fully from the core results in a shifting of the power distribution toward the top of the core and must be accounted for in the design analysis. On the other hand, an optional plan for control rod repositioning that considers margins available in related design parameters can be devised that minimizes the effects of the repositioning for the reload. This paper summarizes a rod repositioning strategy implemented for a recent reload and some calculated power shape results for this strategy and other scenarios

  15. Investigating the Act of Design in Discharge Concept Using PMRI

    Science.gov (United States)

    Lestariningsih; Anwar, Muhammad; Setiawan, Agus Mulyanto

    2015-01-01

    The goal of this research is to investigate the act of design in discharge concept using Pendidikan Matematika Realistik Indonesia (PMRI) approach with Lapindo's Mud phenomenon as a context. Design research was chosen as the method used in this research that consists of three phases, namely preparing for the experiment, teaching experiment, and…

  16. Mechatronical Aided Concept (MAC in Intelligent Transport Vehicles Design

    Directory of Open Access Journals (Sweden)

    Pavel Pavlasek

    2003-01-01

    Full Text Available This article deals with the principles of synergy effect of mechatronical aided concept (MAC to the design of intelligent transport vehicles products applying CA technologies and virtual reality design methods. Also includes presentation of intelligent railway vehicle development.

  17. Preliminary Design Concept for a Reactor-internal CRDM

    International Nuclear Information System (INIS)

    Lee, Jae Seon; Kim, Jong Wook; Kim, Tae Wan; Choi, Suhn; Kim, Keung Koo

    2013-01-01

    A rod ejection accident may cause severer result in SMRs because SMRs have relatively high control rod reactivity worth compared with commercial nuclear reactors. Because this accident would be perfectly excluded by adopting a reactor-internal CRDM (Control Rod Drive Mechanism), many SMRs accept this concept. The first concept was provided by JAERI with the MRX reactor which uses an electric motor with a ball screw driveline. Babcock and Wilcox introduced the concept in an mPower reactor that adopts an electric motor with a roller screw driveline and hydraulic system, and Westinghouse Electric Co. proposes an internal Control Rod Drive in its SMR with an electric motor with a latch mechanism. In addition, several other applications have been reported thus far. The reactor-internal CRDM concept is now widely adopted in many SMR designs, and this concept may also be applied in an evolutionary reactor development. So the preliminary study is conducted based on the SMART CRDM design. A preliminary design concept for a reactor-internal CRDM was proposed and evaluated through an electromagnetic analysis. It was found that there is an optimum design for the motor housing, and the results may contribute to the realization a reactor-internal CRDM for an evolutionary reactor development. More detailed analysis results will be reported later

  18. High Altitude Venus Operations Concept Trajectory Design, Modeling and Simulation

    Science.gov (United States)

    Lugo, Rafael A.; Ozoroski, Thomas A.; Van Norman, John W.; Arney, Dale C.; Dec, John A.; Jones, Christopher A.; Zumwalt, Carlie H.

    2015-01-01

    A trajectory design and analysis that describes aerocapture, entry, descent, and inflation of manned and unmanned High Altitude Venus Operation Concept (HAVOC) lighter-than-air missions is presented. Mission motivation, concept of operations, and notional entry vehicle designs are presented. The initial trajectory design space is analyzed and discussed before investigating specific trajectories that are deemed representative of a feasible Venus mission. Under the project assumptions, while the high-mass crewed mission will require further research into aerodynamic decelerator technology, it was determined that the unmanned robotic mission is feasible using current technology.

  19. Design Process for Integrated Concepts with Responsive Building Elements

    DEFF Research Database (Denmark)

    Aa, Van der A.; Heiselberg, Per

    2008-01-01

    An integrated building concept is a prerequisite to come to an energy efficient building with a good and healthy IAQ indoor comfort. A design process that defines the targets and boundary conditions in the very first stage of the design and guarantees them until the building is finished and used...... is needed. The hard question is however: how to make the right choice of the combination of individual measures from building components and building services elements. Within the framework of IEA-ECBCS Annex 44 research has been conducted about the design process for integrated building concepts...

  20. Techniques for Conducting Effective Concept Design and Design-to-Cost Trade Studies

    Science.gov (United States)

    Di Pietro, David A.

    2015-01-01

    Concept design plays a central role in project success as its product effectively locks the majority of system life cycle cost. Such extraordinary leverage presents a business case for conducting concept design in a credible fashion, particularly for first-of-a-kind systems that advance the state of the art and that have high design uncertainty. A key challenge, however, is to know when credible design convergence has been achieved in such systems. Using a space system example, this paper characterizes the level of convergence needed for concept design in the context of technical and programmatic resource margins available in preliminary design and highlights the importance of design and cost evaluation learning curves in determining credible convergence. It also provides techniques for selecting trade study cases that promote objective concept evaluation, help reveal unknowns, and expedite convergence within the trade space and conveys general practices for conducting effective concept design-to-cost studies.

  1. Seismic responses of a pool-type fast reactor with different core support designs

    International Nuclear Information System (INIS)

    Wu, Ting-shu; Seidensticker, R.W.

    1989-01-01

    In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs

  2. Core/shell PLGA microspheres with controllable in vivo release profile via rational core phase design.

    Science.gov (United States)

    Yu, Meiling; Yao, Qing; Zhang, Yan; Chen, Huilin; He, Haibing; Zhang, Yu; Yin, Tian; Tang, Xing; Xu, Hui

    2018-02-27

    the microspheres prepared by various methods were mainly controlled by either the porosity inside the microspheres or the degradation of materials, which could, therefore, lead to different release behaviours. This results indicated great potential of the PLGA microsphere formulation as an injectable depot for controllable in vivo release profile via rational core phase design. Core/shell microspheres fabricated by modified double emulsification-solvent evaporation methods, with various inner phases, to obtain high loading drugs system, as well as appropriate release behaviours. Accordingly, control in vivo release profile via rational core phase design.

  3. Combined raman spectrometer/laser-induced breakdown spectrometer design concept

    Science.gov (United States)

    Bazalgette Courrèges-Lacoste, Gregory; Ahlers, Berit; Boslooper, Erik; Rull-Perez, Fernando; Maurice, Sylvestre

    2017-11-01

    Amongst the different instruments that have been preselected to be on-board the Pasteur payload on ExoMars is the Raman/ Laser Induced Breakdown Spectroscopy (LIBS) instrument. Raman spectroscopy and LIBS will be integrated into a single instrument sharing many hardware commonalities. An international team under the lead of TNO has been gathered to produce a design concept for a combined Raman Spectrometer/ LIBS Elegant Bread-Board (EBB). The instrument is based on a specifically designed extremely compact spectrometer with high resolution over a large wavelength range, suitable for both Raman spectroscopy and LIBS measurements. Low mass, size and resources are the main drivers of the instrument's design concept. The proposed design concept, realization and testing programme for the combined Raman/ LIBS EBB is presented as well as background information on Raman and LIBS.

  4. Bypass Flow and Hot Spot Analysis for PMR200 Block-Core Design with Core Restraint Mechanism

    International Nuclear Information System (INIS)

    Lim, Hong Sik; Kim, Min Hwan

    2009-01-01

    The accurate prediction of local hot spot during normal operation is important to ensure core thermal margin in a very high temperature gas-cooled reactor because of production of its high temperature output. The active cooling of the reactor core determining local hot spot is strongly affected by core bypass flows through the inter-column gaps between graphite blocks and the cross gaps between two stacked fuel blocks. The bypass gap sizes vary during core life cycle by the thermal expansion at the elevated temperature and the shrinkage/swelling by fast neutron irradiation. This study is to investigate the impacts of the variation of bypass gaps during core life cycle as well as core restraint mechanism on the amount of bypass flow and thus maximum fuel temperature. The core thermo fluid analysis is performed using the GAMMA+ code for the PMR200 block-core design. For the analysis not only are some modeling features, developed for solid conduction and bypass flow, are implemented into the GAMMA+ code but also non-uniform bypass gap distribution taken from a tool calculating the thermal expansion and the shrinkage/swell of graphite during core life cycle under the design options with and without core restraint mechanism is used

  5. Neutronic and mechanical design of the reactor core of the Opus system

    Energy Technology Data Exchange (ETDEWEB)

    Raepsaet, X.; Pascal, S. [CEA Saclay, Dept. Modelisation de Systemes et Structures (DEN/DM2S), 91 - Gif sur Yvette (France)

    2007-07-01

    Since a few years now, Cea decided to maintain a waking state in its space nuclear activities by carrying out some conceptual studies of embarked nuclear power systems in the range of 100-500 kWe. Results stemming from these ongoing studies are gathered in the project OPUS -Optimized Propulsion Unit System-. This nuclear power system relies on a fast gas-cooled reactor concept coupled either to a Brayton cycle or to a more ambitious energy conversion system using a Hirn cycle to dramatically reduce the size of the radiator. The OPUS reactor core consists of an arrangement of enriched graphite elements of hexagonal cross-section. Their length is equal to the core diameter (48 cm). Coated fuel particles containing enriched (93%) uranium are embedded in these fuel elements. Each fuel element is designed with a centered axial channel through which flows the working fluid: a mixture of helium and xenon gas. This reactor is expected to have an operating life of over 2000 days at full power. In fact the main questions remain on the fuel element manufacturing and on the mechanical design (type and size of particles, packing fraction in the matrix, final core diameter and mass). Especially, the nuclear reactor has been defined considering the possible synergies with the next generation of terrestrial nuclear reactor (International Generation IV Forum). Based on relatively short-term technologies, the same reactor is designed to cover a wide range of power: 100 to 500 kWe without core design modification. The final reactor design presented in this paper is the result of a coupled analysis between the thermomechanical and the neutronic aspects.

  6. New concepts for controlled fusion reactor blanket design

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Avci, H.; El-Maghrabi, M.

    1975-01-01

    Several new concepts for fusion reactor blanket design based on the idea of shifting, or tailoring, the neutron spectrum incident on the first structural wall are presented. The spectral shifter is a nonstructural element which can be made of graphite, silicon carbide, or three dimensionally woven carbon fibers (and containing other materials as appropriate) placed between the neutron source and the first structural wall. The softened neutron spectrum incident on the structural components leads to lower gas production and atom displacement rates than in more standard fusion blanket designs. In turn, this results in longer anticipated lifetimes for the structural materials and can significantly reduce radioactivity and afterheat levels. In addition, the neutron spectrum in the first structural wall can be made to approach the flux shape in fast breeder reactors. Such spectral softening means that existing radiation facilities may be more profitably used to provide relevant materials radiation damage data for the structural materials in these fusion blanket designs. This general class of blanket concepts are referred to as internal spectral shifter and energy converter, or ISSEC concepts. These specific design concepts fall into three main categories: ISSEC/EB concepts based on utilizing existing designs which breed tritium behind the first structural wall; ISSEC/IB concepts based on breeding tritium inside the first vacuum wall; and ISSEC/Bu concepts based on using boron, carbon, and perhaps, beryllium to obtain an energy multiplier and converter design that does not attempt to breed tritium or utilize lithium. The detailed analyses relate specifically to the nuclear performance of ISSEC systems and to a discussion of materials radiation damage problems in the structural material.(U.S.)

  7. Systems analysis and futuristic designs of advanced biofuel factory concepts.

    Energy Technology Data Exchange (ETDEWEB)

    Chianelli, Russ; Leathers, James; Thoma, Steven George; Celina, Mathias C.; Gupta, Vipin P.

    2007-10-01

    The U.S. is addicted to petroleum--a dependency that periodically shocks the economy, compromises national security, and adversely affects the environment. If liquid fuels remain the main energy source for U.S. transportation for the foreseeable future, the system solution is the production of new liquid fuels that can directly displace diesel and gasoline. This study focuses on advanced concepts for biofuel factory production, describing three design concepts: biopetroleum, biodiesel, and higher alcohols. A general schematic is illustrated for each concept with technical description and analysis for each factory design. Looking beyond current biofuel pursuits by industry, this study explores unconventional feedstocks (e.g., extremophiles), out-of-favor reaction processes (e.g., radiation-induced catalytic cracking), and production of new fuel sources traditionally deemed undesirable (e.g., fusel oils). These concepts lay the foundation and path for future basic science and applied engineering to displace petroleum as a transportation energy source for good.

  8. Reduced Graphene Oxide Joins Graphene Oxide to Teach Undergraduate Students Core Chemistry and Nanotechnology Concepts

    Science.gov (United States)

    Kondratowicz, Izabela; Nadolska, Malgorzata; Z?elechowska, Kamila

    2018-01-01

    Novel carbon nanomaterials such as reduced graphene oxide (rGO) and graphene oxide (GO) can be easily incorporated into the undergraduate curriculum to discuss basic chemistry and nanotechnology concepts. This paper describes a laboratory experiment designed to study the differences between GO and rGO regarding their physicochemical properties…

  9. A concept of passive safety pressurized water reactor system with inherent matching nature of core heat generation and heat removal

    International Nuclear Information System (INIS)

    Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

    1995-01-01

    The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of 'safety class' for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation. (author)

  10. Exploring Many-Core Design Templates for FPGAs and ASICs

    Directory of Open Access Journals (Sweden)

    Ilia Lebedev

    2012-01-01

    Full Text Available We present a highly productive approach to hardware design based on a many-core microarchitectural template used to implement compute-bound applications expressed in a high-level data-parallel language such as OpenCL. The template is customized on a per-application basis via a range of high-level parameters such as the interconnect topology or processing element architecture. The key benefits of this approach are that it (i allows programmers to express parallelism through an API defined in a high-level programming language, (ii supports coarse-grained multithreading and fine-grained threading while permitting bit-level resource control, and (iii reduces the effort required to repurpose the system for different algorithms or different applications. We compare template-driven design to both full-custom and programmable approaches by studying implementations of a compute-bound data-parallel Bayesian graph inference algorithm across several candidate platforms. Specifically, we examine a range of template-based implementations on both FPGA and ASIC platforms and compare each against full custom designs. Throughout this study, we use a general-purpose graphics processing unit (GPGPU implementation as a performance and area baseline. We show that our approach, similar in productivity to programmable approaches such as GPGPU applications, yields implementations with performance approaching that of full-custom designs on both FPGA and ASIC platforms.

  11. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes.

  12. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    International Nuclear Information System (INIS)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun

    2016-01-01

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes

  13. Concept design on RH maintenance of CFETR Tokamak reactor

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Songtao; Wan, Yuanxi; Li, Jiangang; Ye, Minyou; Zheng, Jinxing; Cheng, Yong; Zhao, Wenlong; Wei, Jianghua

    2014-01-01

    Highlights: •We discussed the concept design of the RH maintenance system based on the main design work of the key components for CFETR. •The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. •The technical problems encountered in the design process were discussed. •The present concept design of remote maintenance system in this paper can meet the physical and engineering requirement of CFETR. -- Abstract: CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed

  14. Human-centered incubator: beyond a design concept

    OpenAIRE

    Goossens, R H M; Willemsen, H

    2013-01-01

    We read with interest the paper by Ferris and Shepley1 on a human-centered design project with university students on neonatal incubators. It is interesting to see that in the design solutions and concepts as presented by Ferris and Shepley,1 human-centered design played an important role. In 2005, a master thesis project was carried out in the Delft University of Technology, following a similar human-centered design approach.2, 3 In that design project we also addressed the noise level insid...

  15. Book Review. Inger Marie Lid: Universal design. Core values​​, knowledge, and practices

    Directory of Open Access Journals (Sweden)

    Sigmund Asmervik

    2013-12-01

    Full Text Available Sigmund Asmervik, dr.ing. Professor emeritus, Norwegian University of Life Sciences (UMB, has reviewed Inger Marie Lid’s book Universell utforming: Verdigrunnlag, kunnskap og praksis (Universal design: Core values​​, knowledge, and practices. He states that this book provides very good suggestions about topics relating to issues such as how people should be able to develop in the community with others and how to look after issues relating to dignity and bodily vulnerability. This expands and challenges the discourse around the concept of universal design, and is the book's most important contribution. As a textbook aimed at students of architecture and engineering, occupational therapy and physiotherapy studies, and otherwise to anyone who is engaged in universal design in practice, as it says on the publisher's website, it has some significant challenges.

  16. Gas core reactor power plants designed for low proliferation potential

    International Nuclear Information System (INIS)

    Lowry, L.L.

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

  17. Ideas to Design an in situ Diamond Drilling Core Splitter within Soft ...

    African Journals Online (AJOL)

    Michael O. Mensah

    2015-12-02

    Dec 2, 2015 ... the wireline system of core barrel assembly and the device used in splitting of core ... Keywords: Design, In situ, Diamond drilling, Core splitter, Wireline system .... This is the most complex part of the core barrel and has many.

  18. Preliminary design concept of HYPER cooling system using Pb-Bi coolant

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang H

    2001-09-01

    The present study focuses on providing the basic concept of HYPER's cooling system based on simple and fundamental calculations. The system operating temperature was preliminarily determined as 340/510 .deg. C. The total system flow rate of HYPER is {approx} 40,000kg/sec and the flow velocity in the core is preliminarily designed to be {approx}1.5 m/sec. For hot conditions of HYPER core, the simple analytic calculation predicted that the maximum temperature of the cladding outer surface is 634 .deg. C, which is below the design limit, 650 .deg. C. However, the SLTHEN code modified for HYPER's subchannel analysis predicted that the maximum temperature of the cladding outer surface in the same conditions is higher than the design limit by 4.7 .deg. C. The comparison with the results of the analytic model and additional sensitivity calculations showed that the modified SLTHEN code can reasonably simulate the heat transfer between subchannels of the HYPER core and be used effectively for thermal hydraulic design of the HYPER core in conceptual design stage. A forced circulation is inevitable during a full power condition since natural circulation is not sufficient to cool the core with reasonable system pressure drop and reasonable system height. However, a natural circulation can be an excellent method for decay heat removal when the height difference between the core and the heat exchanger is above 10 m. In order to avoid high pressure loads on the vessel, loop configuration was chosen. The simplification of cooling system and high system efficiency were attained by removing independent target cooling system and intermediate heat transport system. A superheated rankle cycle was chosen since it is technically matured and its thermal efficiency is reasonably high.

  19. Concept Design of Movable Beam of Hydraulic Press

    Directory of Open Access Journals (Sweden)

    Li Yancong

    2017-01-01

    Full Text Available The hydraulic press movable beam is one of the key components of the hydraulic press; its design quality impacts the accuracy of the workpiece that the press suppressed. In this paper, first, with maximum deflection and material strength as constraints, mechanical model of the movable beam is established; next, the concept design model of the moveable beam structure is established; the relationship among the force of the side cylinder, the thickness of the inclined plate, outer plate is established also. Taking movable beam of the 100MN type THP10-10000 isothermal forging hydraulic press as an example, the conceptual design result is given. This concept design method mentoned in the paper has general meaning and can apply to other similar product design.

  20. CONTEMPORARY DESIGN CONCEPTS IN CAMOUFLAGE: UNIVERSAL VERSUS SPECIALIZED

    Directory of Open Access Journals (Sweden)

    TACHEV Momchil

    2015-06-01

    Full Text Available In present days there is a strong influence of the concept incorporating increased use of low intensive colours and small “digitalized” form structures for modern army camouflage designs. It was proclaimed as a revolutionary “universal” design line. It was supposed to be superior to others patterns designs and to be optimal for most types of environment. Meanwhile, some arm forces insist on developing more limited as terrain range camouflage design but with better efficiency and specialization. These two concepts in design reflect the political and the military philosophies of the countries they represent. However, at the end for the soldiers at the battlefield it is a matter of survival.

  1. Fast reactor calculational route for Pu burning core design

    Energy Technology Data Exchange (ETDEWEB)

    Hunter, S. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    This document provides a description of a calculational route, used in the Reactor Physics Research Section for sensitivity studies and initial design optimization calculations for fast reactor cores. The main purpose in producing this document was to provide a description of and user guides to the calculational methods, in English, as an aid to any future user of the calculational route who is (like the author) handicapped by a lack of literacy in Japanese. The document also provides for all users a compilation of information on the various parts of the calculational route, all in a single reference. In using the calculational route (to model Pu burning reactors) the author identified a number of areas where an improvement in the modelling of the standard calculational route was warranted. The document includes comments on and explanations of the modelling assumptions in the various calculations. Practical information on the use of the calculational route and the computer systems is also given. (J.P.N.)

  2. Design studies of back up cores for the experimental multi-purpose VHTR, (1)

    International Nuclear Information System (INIS)

    Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu

    1982-09-01

    For the Experimental Multi-Purpose Very High Temperature Reactor, design studies have been made of two backup cores loaded with new type fuel elements. The purpose is to improve core operational characteristics of the standard design core (Mark-III core) consisting of pin-in-block type fuel element having externally cooled hollow fuel rods. The first backup core (semi-pin fuel core) is composed of fuel elements with internally cooled fuel pins, and the second core (multihole fuel core) is composed of multihole fuel elements, which can be adopted for the experimental VHTR as the substitution of the standard Mark-III fuel element. Either of the cores has 73 fuel columns and 4 m height. The arrangement of active core and reactor internal structure is same as that in the standard design core. These backup cores meet almost all design requirements of the VHTR and increase the margins for some important design items in comparison with the standard core (Mark-III core). This report describes the overall characteristics of nuclear, thermal-hydraulic, fuel and safety, and structural consideration for these cores. (author)

  3. Pre-conceptual core design of SCWR with annular fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Chuanqi [Key Laboratory of Thermo-Fluid Science and Engineering of MOE, School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); Cao, Liangzhi, E-mail: caolz@mail.xjtu.edu.cn [Key Laboratory of Thermo-Fluid Science and Engineering of MOE, School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); Wu, Hongchun; Zheng, Youqi [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China)

    2014-02-15

    Highlights: • Annular fuel with both internal and external cooling is used in supercritical light water reactor (SCWR). • The geometry of the annular fuel has been optimized to achieve better performance for the SCWR. • Based on the annular fuel assembly, an equilibrium core has been designed. • The results show that the equilibrium core has satisfied all the objectives and design criteria. - Abstract: The new design of supercritical light water reactor was proposed using annular fuel assemblies. Annular fuel consists of several concentric rings. Feed water flows through the center and outside of the fuel to give both internal and external cooling. Thanks to this feature, the fuel center temperature and the cladding temperature can be reduced and high power density can be achieved. The water flowing through the center also provides moderation, so there is no need for extra water rods in the assembly. The power distribution can be easily flattened by use of this design. The geometry of the annular fuel has been optimized to achieve better performance for the SCWR. There are 19 fuel pins in an assembly. Burnable poison is utilized to reduce the initial excess reactivity. The fuel reloading pattern and water flow scheme were optimized to achieve more uniform power distribution and lower cladding temperature. An equilibrium core has been designed and analyzed using three dimensional neutronics and thermal-hydraulics coupling calculations. The void reactivity, Doppler coefficient and cold shut down margin were calculated for safety consideration. The present results show that this concept is a promising design for the SCWR.

  4. Inherent safe design of advanced high temperature reactors - concepts for future nuclear power plants

    International Nuclear Information System (INIS)

    Hodzic, A.; Kugeler, K.

    1997-01-01

    This paper discusses the applicable solutions for a commercial size High Temperature Reactor (HTR) with inherent safety features. It describes the possible realization using an advanced concept which combines newly proposed design characteristics with some well known and proven HTR inherent safety features. The use of the HTR technology offers the conceivably best solution to meet the legal criteria, recently stated in Germany, for the future reactor generation. Both systems, block and pebble bed ,reactor, could be under certain design conditions self regulating in terms of core nuclear heat, mechanical stability and the environmental transfer. 23 refs., 7 figs

  5. Conceptual design of laser fusion reactor, SENRI-I - 1. concept and system design

    International Nuclear Information System (INIS)

    Ido, S.; Naki, S.; Norimatsu, T.

    1981-01-01

    Design features of a laser fusion reactor concept SENRI-I and new concepts are reviewed and discussed. The unique feature is the utilization of a magnetic field to guide and control the inner liquid Li flow. Basic requirements and typical parameters used in the design are presented. Items to be discussed are constitution of the system, performance of liquid Li flow, neutronics, thermo-electric cycle, fuel cycle and new concepts

  6. Concept designs for NASA's Solar Electric Propulsion Technology Demonstration Mission

    Science.gov (United States)

    Mcguire, Melissa L.; Hack, Kurt J.; Manzella, David H.; Herman, Daniel A.

    2014-01-01

    Multiple Solar Electric Propulsion Technology Demonstration Mission were developed to assess vehicle performance and estimated mission cost. Concepts ranged from a 10,000 kilogram spacecraft capable of delivering 4000 kilogram of payload to one of the Earth Moon Lagrange points in support of future human-crewed outposts to a 180 kilogram spacecraft capable of performing an asteroid rendezvous mission after launched to a geostationary transfer orbit as a secondary payload. Low-cost and maximum Delta-V capability variants of a spacecraft concept based on utilizing a secondary payload adapter as the primary bus structure were developed as were concepts designed to be co-manifested with another spacecraft on a single launch vehicle. Each of the Solar Electric Propulsion Technology Demonstration Mission concepts developed included an estimated spacecraft cost. These data suggest estimated spacecraft costs of $200 million - $300 million if 30 kilowatt-class solar arrays and the corresponding electric propulsion system currently under development are used as the basis for sizing the mission concept regardless of launch vehicle costs. The most affordable mission concept developed based on subscale variants of the advanced solar arrays and electric propulsion technology currently under development by the NASA Space Technology Mission Directorate has an estimated cost of $50M and could provide a Delta-V capability comparable to much larger spacecraft concepts.

  7. Design Core Commonalities: A Study of the College of Design at Iowa State University

    Science.gov (United States)

    Venes, Jane

    2015-01-01

    This comprehensive study asks what a group of rather diverse disciplines have in common. It involves a cross-disciplinary examination of an entire college, the College of Design at Iowa State University. This research was intended to provide a sense of direction in developing and assessing possible core content. The reasoning was that material…

  8. Concept of modular flexure-based mechanisms for ultra-high precision robot design

    Directory of Open Access Journals (Sweden)

    M. Richard

    2011-05-01

    Full Text Available This paper introduces a new concept of modular flexure-based mechanisms to design industrial ultra-high precision robots, which aims at significantly reducing both the complexity of their design and their development time. This modular concept can be considered as a robotic Lego, where a finite number of building bricks is used to quickly build a high-precision robot. The core of the concept is the transformation of a 3-D design problem into several 2-D ones, which are simpler and well-mastered. This paper will first briefly present the theoretical bases of this methodology and the requirements of both types of building bricks: the active and the passive bricks. The section dedicated to the design of the active bricks will detail the current research directions, mainly the maximisation of the strokes and the development of an actuation sub-brick. As for the passive bricks, some examples will be presented, and a discussion regarding the establishment of a mechanical solution catalogue will conclude the section. Last, this modular concept will be illustrated with a practical example, consisting in the design of a 5-degree of freedom ultra-high precision robot.

  9. Design for All in Scandinavia - a strong concept.

    Science.gov (United States)

    Bendixen, Karin; Benktzon, Maria

    2015-01-01

    Design for All is more than an appealing point of view. It is a concept that offers a set of challenges capable of generating innovation and giving design added value and weight. In the Scandinavian tradition, the concept has developed from a purely social dimension to a design topic that is discussed both in terms of its business potential and in relation to Corporate Social Responsibility, CSR. This article gives a State of the Art of the development of Design for All in the Scandinavian countries: Denmark, Norway, Sweden and Finland during the past 15 years, beginning with a common review and joint Scandinavian projects, followed by an overall review country by country which include selected case studies over the past 15 years. Copyright © 2013 Elsevier Ltd and The Ergonomics Society. All rights reserved.

  10. Development of UCMS for Analysis of Designed and Measured Core Power Distribution

    International Nuclear Information System (INIS)

    Moon, Sang Rae; Hong, Sun Kwan; Yang, Sung Tae

    2009-01-01

    In this study, reactor core loading patterns were determined by calculating and verifying the factors affecting peak power and important core safety variables were reconciled with their design criteria using a newly designed unified core management system. Core loading patterns are designed for quadrant cores under the assumption that the power distribution of the reactor core is the same among symmetric fuel assemblies within the core. Actual core power distributions measured during core operation may differ slightly from their designed data. Reactor engineers monitor these differences between the designed and measured data by performing a surveillance procedure every month according to the technical specification requirements. It is difficult to monitor overall power distribution behavior throughout the assemblies using the current procedure because it requires the reactor engineer to compare the designed data with only the maximum value of the power peaking factor and the relative power density. It is necessary to enhance this procedure to check the primary variables such as core power distribution, because long cycle operation, high burnup, power up-rate, and improved fuel can change the environment in the core. To achieve this goal, a web-based Unified Core Management System (UCMS) was developed. To build the UCMS, a database system was established using reactor design data such as that in the Nuclear Design Report (NDR) and automated core analysis codes for all light water reactor power plants. The UCMS is designed to help reactor engineers to monitor important core variables and core safety margins by comparing the measured core power distribution with designed data for each fuel assembly during the cycle operation in nuclear power plants

  11. 78 FR 32988 - Core Principles and Other Requirements for Designated Contract Markets; Correction

    Science.gov (United States)

    2013-06-03

    ... COMMODITY FUTURES TRADING COMMISSION 17 CFR Part 38 RIN 3038-AD09 Core Principles and Other... regarding Core Principles and Other Requirements for Designated Contract Markets by inserting a missing... regarding Core Principles and Other Requirements for Designated Contract Markets (77 FR 36612, June 19, 2012...

  12. From corporate control concept to ERP system design

    NARCIS (Netherlands)

    Kerssens-van Drongelen, I.C.

    2003-01-01

    As Goold and Campbell (2002) rightly observed in their recent article, currently not many organizations have purposefully designed their organization set-up and the other elements of their corporate control concept such as the management style, the responsibility structure, and the format and use of

  13. An integral design concept for ecological self-compacting concrete

    NARCIS (Netherlands)

    Hunger, M.

    2010-01-01

    This Thesis addresses an alternative design concept for Self-Compacting Concrete (SCC). SCC is a special type of concrete with superior workability, which flows and compacts in all corners of a formwork just by the influence of gravity. Introduced to the concrete world in the late 1980s, SCC has

  14. The Concept of Design: Are We off Track?

    Science.gov (United States)

    2012-06-15

    Andrews, Henry Mintzberg, Ludwig von Bertalanffy, Thomas Kuhn, Horst Rittel, Melvin Webber, Dietrich Dörner, and Peter Senge listed below cement a...perspective. The terms, concepts, and intellectual accessories presented in this chapter cement a deeper appreciation and understanding of design as a...thinking. Creative thinking involves generating something new or original. It involves the skills of flexibility, originality, fluency , elaboration

  15. Back up core designs for the experimental multi-purpose VHTR

    International Nuclear Information System (INIS)

    Aochi, Tetsuo; Yasuno, Takehiko; Miyamoto, Yoshiaki; Shindo, Ryuichi; Ikushima, Takeshi

    1979-02-01

    For the Experimental Multi-Purpose Very High Temperature Reactor (thermal power 50 MW and reactor outlet helium temperature 1000 0 C), design studies have been made of two backup cores loaded with new-type fuel elements. The purpose is to improve core operational characteristics, especially in thermohydraulics, of the reference design core consisting of pin-in-block type fuel elements having externally cooled hollow fuel rods. In this report are described the design principles and the analyses made of nuclear, thermal and hydraulic, fuel, and safety performances to determine the backup fuel and core design parameters. The first backup core (SP fuel core) is composed of fuel elements with internally cooled fuel rods (semi-pin), 36 rods in each standard element and 18 rods in each control element. The second backup core (MH fuel core) is composed of multihole fuel elements. 102 fuel and 54 coolant holes in each standard element and 30 fuel and 18 coolant holes in each control element. Either of the cores has 73 fuel columns 4 m high; the arrangement of active core and reactor internal structures is the same as that in the reference design. The backup cores meet nearly all design requirements of the VHTR, permitting the rated power operation with coolant Reynolds number of over 10,000 in the SP core and over 6,000 in the MH core. (author)

  16. A novel concept of fault current limiter based on saturable core in high voltage DC transmission system

    Science.gov (United States)

    Yuan, Jiaxin; Zhou, Hang; Gan, Pengcheng; Zhong, Yongheng; Gao, Yanhui; Muramatsu, Kazuhiro; Du, Zhiye; Chen, Baichao

    2018-05-01

    To develop mechanical circuit breaker in high voltage direct current (HVDC) system, a fault current limiter is required. Traditional method to limit DC fault current is to use superconducting technology or power electronic devices, which is quite difficult to be brought to practical use under high voltage circumstances. In this paper, a novel concept of high voltage DC transmission system fault current limiter (DCSFCL) based on saturable core was proposed. In the DCSFCL, the permanent magnets (PM) are added on both up and down side of the core to generate reverse magnetic flux that offset the magnetic flux generated by DC current and make the DC winding present a variable inductance to the DC system. In normal state, DCSFCL works as a smoothing reactor and its inductance is within the scope of the design requirements. When a fault occurs, the inductance of DCSFCL rises immediately and limits the steepness of the fault current. Magnetic field simulations were carried out, showing that compared with conventional smoothing reactor, DCSFCL can decrease the high steepness of DC fault current by 17% in less than 10ms, which verifies the feasibility and effectiveness of this method.

  17. Analytic function expansion nodal method for nuclear reactor core design

    International Nuclear Information System (INIS)

    Noh, Hae Man

    1995-02-01

    than the analytic function. The second variation of the AFEN method we developed is the AFEN/PEN hybrid method. This method is designed especially for the multigroup reactor analysis. This hybrid method solves the diffusion equations for the fast energy groups by the PEN method, and those for the thermal energy groups by the AFEN method. This method is based on the observation that the fast group neutron flux distributions are generally so smooth that they can be approximated by a high-order polynomial and that, on the other hand, the thermal fluxes require the analytic function expansion for the representation of their strong gradients near the interface between assemblies having different neutronic properties. The results of benchmark problems on which this method was tested indicate that performance of the hybrid method is much better than that of the PEN method and is nearly the same to that of the AFEN method. In order for the AFEN method and its variations to be used in analyzing the neutron behavior in an actual reactor core, we also developed a new burnup correction model to reduce the errors in nodal flux distributions induced by the intranodal burnup gradients. It is essential for the nodal methods to maintain their accuracy in fuel depletion analysis. The burnup correction model developed in this study homogenizes equivalently the node with the burnup-induced cross section variations into the homogeneous node with the equivalent parameters such as the flux-volume-weighted constant cross sections and the discontinuity factors. The results of a benchmark problem show that this model eliminates almost all the errors in the nodal unknowns which are induced by the intranodal burnup gradients

  18. Research on green building design based on ecological concept

    Directory of Open Access Journals (Sweden)

    Zhang Ping Qing

    2016-01-01

    Full Text Available At present, the protection of the ecological environment and the promotion of green building has been recognized and widely promoted.With the rapid development of the construction industry, Architecture design will inevitably require the resentation of its unique form and charm to reflect the ecological concept and ecological culture, because of the unique nature of the art and the particularity of the environment. To establish the ecological concept of green building design and vigorously develop the green green building has a complementary role to alleviate the pressure on resources,and to speed up the eco city planning design, and to realize the sustainable development of the city, and to protect the urban ecological environmental.

  19. Design concept of HYPER (HYbrid Power Extraction Reactor)

    International Nuclear Information System (INIS)

    Park, Won S.; Song, Tae Y.; Yu, Dong H.; Kim, Chang H.

    1999-01-01

    Korea Atomic Energy Research Institute (KAERI) has been performing accelerator driven system related research and development called HYPER for the transmutation of nuclear waste and energy production through the transmutation process. Some major design features of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with a minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Pb-Bi is adopted as a coolant and spallation target material. 1 GeV 13 mA proton beam is designed to be provided for HYPER. HYPER is to transmute about 380 kg of TRU a year and produce 1000 MWth power. The support ratio of HYPER for LWR units producing the same power is believed to be 5 to 6. (author)

  20. Core design for use with precision composite reflectors

    Science.gov (United States)

    Porter, Christopher C. (Inventor); Jacoy, Paul J. (Inventor); Schmitigal, Wesley P. (Inventor)

    1992-01-01

    A uniformly flexible core, and method for manufacturing the same, is disclosed for use between the face plates of a sandwich structure. The core is made of a plurality of thin corrugated strips, the corrugations being defined by a plurality of peaks and valleys connected to one another by a plurality of diagonal risers. The corrugated strips are orthogonally criss-crossed to form the core. The core is particularly suitable for use with high accuracy spherically curved sandwich structures because undesirable stresses in the curved face plates are minimized due to the uniform flexibility characteristics of the core in both the X and Y directions. The core is self venting because of the open geometry of the corrugations. The core can be made from any suitable composite, metal, or polymer. Thermal expansion problems in sandwich structures may be minimized by making the core from the same composite materials that are selected in the manufacture of the curved face plates because of their low coefficients of thermal expansion. Where the strips are made of a composite material, the core may be constructed by first cutting an already cured corrugated sheet into a plurality of corrugated strips and then secondarily bonding the strips to one another or, alternatively, by lying a plurality of uncured strips orthogonally over one another in a suitable jig and then curing and bonding the entire plurality of strips to one another in a single operation.

  1. Study on core flow distribution of the reference core design Mark-III of experimental multi-purpose VHTR

    International Nuclear Information System (INIS)

    Satoh, Sadao; Arai, Taketoshi; Miyamoto, Yoshiaki; Hirano, Mitsumasa

    1977-01-01

    Concerning the coolant flow distribution between fuel channels and other flow paths in the core, designated as Reference Core Mark-III of the Multi-purpose Experimental Very High Temperature Reactor, thermal analysis has been made of the control rods and other steel structures around the core to find the coolant flow rates (bypass flow) necessary to cool them to their safe operating temperatures. Calculations showed that adequate cooling could be achieved in the Mark-III Core by the bypass flow of 8% of the total reactor coolant flow, 4% each for the control-rod channels and for other structures. The thermal and coolant flow design bases, including the assumption of a 10% bypass flow, were thus confirmed to first approximation. (auth.)

  2. The conceptual basis for interprofessional collaboration: core concepts and theoretical frameworks.

    Science.gov (United States)

    D'Amour, Danielle; Ferrada-Videla, Marcela; San Martin Rodriguez, Leticia; Beaulieu, Marie-Dominique

    2005-05-01

    Interprofessional collaboration is a key factor in initiatives designed to increase the effectiveness of health services currently offered to the public. It is important that the concept of collaboration be well understood, because although the increasingly complex health problems faced by health professionals are creating more interdependencies among them, we still have limited knowledge of the complexity of interprofessional relationships. The goal of this literature review was to identify conceptual frameworks that could improve our understanding of this important aspect of health organizations. To this end, we have identified and taken into consideration: (A) the various definitions proposed in the literature and the various concepts associated with collaboration, and (B) the various theoretical frameworks of collaboration. Our results demonstrate that: (1) the concept of collaboration is commonly defined through five underlying concepts: sharing, partnership, power, interdependency and process; (2) the most complete models of collaboration seem to be those based on a strong theoretical background, either in organizational theory or in organizational sociology and on empirical data; (3) there is a significant amount of diversity in the way the various authors conceptualized collaboration and in the factors influencing collaboration; (4) these frameworks do not establish clear links between the elements in the models and the outputs; and (5) the literature does not provide a serious attempt to determine how patients could be integrated into the health care team, despite the fact that patients are recognized as the ultimate justification for providing collaborative care.

  3. New concept of composite strengthening in Co-Re based alloys for high temperature applications in gas turbines

    Energy Technology Data Exchange (ETDEWEB)

    Mukherji, D.; Roesler, J.; Fricke, T.; Schmitz, F. [Technische Univ. Braunschweig (DE). Inst. fuer Werkstoffkunde (IfW); Piegert, S. [Siemens AG, Berlin (DE). Energy Sector (F PR GT EN)

    2010-07-01

    High temperature material development is mainly driven by gas turbine needs. Today, Ni-based superalloys are the dominant material class in the hot section of turbines. Material development will continue to push the maximum service temperature of Ni-superalloys upwards. However, this approach has a fundamental limit and can not be sustained indefinitely, as the Ni-superalloys are already used very close to their melting point. Within the frame work of a DFG Forschergruppe program (FOR 727) - ''Beyond Ni-base Superalloys'' - Co-Re based alloys are being developed as a new generation of high temperature materials that can be used at +100 C above single crystal Ni-superalloys. Along with other strengthening concepts, hardening by second phase is explored to develop a two phase composite alloy. With quaternary Co-Re-Cr-Ni alloys we demonstrate this development concept, where Co{sub 2}Re{sub 3}-type {sigma} phase is used in a novel way as the hardening phase. Thermodynamic calculation was used for designing model alloy compositions. (orig.)

  4. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

    2014-10-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

  5. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    International Nuclear Information System (INIS)

    Lazaro, A.; Schikorr, M.; Mikityuk, K.; Ammirabile, L.; Bandini, G.; Darmet, G.; Schmitt, D.; Dufour, Ph.; Tosello, A.; Gallego, E.; Jimenez, G.; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D.; Stempniewicz, M.

    2014-01-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs

  6. A heuristic application of critical power ratio to pressurized water reactor core design

    International Nuclear Information System (INIS)

    Ahn, Seung Hoon; Jeun, Gyoo Dong

    2002-01-01

    The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR correlation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large

  7. Neutronic design of a traveling wave reactor core

    International Nuclear Information System (INIS)

    Lopez S, R. C.; Francois L, J. L.

    2010-10-01

    The traveling wave reactor is an innovative kind of fast breeder reactor, capable of operate for decades without refueling and whose operation requires only a small amount of enriched fuel for the ignition. Also, one of its advantages is its versatility; it can be designed as small modules of about 100 M We or large scale units of 1000 M We. In this paper the behaviour of the traveling wave reactor core is studied in order to determine whether the traveling breeding/burning wave moves (as theoretically predicted) or not. To achieve this, we consider a two pieces cylinder, the first one, the ignition zone, containing highly enriched fuel and the second, the breeding zone, which is the larger, containing natural or depleted uranium or thorium. We consider that both zones are homogeneous mixtures of fuel, sodium as coolant and iron as structural material. We also include a reflector material outside the cylinder to reduce the neutron leakages. Simulations were run with MCNPX version 2.6 code. We observed that the wave does move as time passes as predicted by theory, and reactor remains supercritical in the time we have simulated (3000 days). Also, we found that thorium does not perform as well as uranium for breeding in this type of reactor. Further test with different reflectors are planned for both U-Pu and Th-U fuel cycles. (Author)

  8. Improving the calculated core stability by the core nuclear design optimization

    International Nuclear Information System (INIS)

    Partanen, P.

    1995-01-01

    Three different equilibrium core loadings for TVO II reactor have been generated in order to improve the core stability properties at uprated power level. The reactor thermal power is assumed to be uprated from 2160 MW th to 2500 MW th , which moves the operating point after a rapid pump rundown where the core stability has been calculated from 1340 MW th and 3200 kg/s to 1675 MW th and 4000 kg/s. The core has been refuelled with ABB Atom Svea-100 -fuel, which has 3,64% w/o U-235 average enrichment in the highly enriched zone. PHOENIX lattice code has been used to provide the homogenized nuclear constants. POLCA4 static core simulator has been used for core loadings and cycle simulations and RAMONA-3B program for simulating the dynamic response to the disturbance for which the stability behaviour has been evaluated. The core decay ratio has been successfully reduced from 0,83 to 0,55 mainly by reducing the power peaking factors. (orig.) (7 figs., 1 tab.)

  9. An innovative fuel design concept for improved Light Water Reactor performance and safety

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Connell, R.G.

    1993-01-01

    The primary goal of this research is to develop a new fuel design which will have improved thermal/mechanical performance characteristics greatly superior to current thermal and mechanical design performance. The mechanical/thermal constraints define the lifetime of the fuel, the maximum power at which the fuel can be operated, the probability of fuel failure over core lifetime, and the integrity of a core during a transient excursion. The thermal/mechanical limits act to degrade fuel integrity when they are violated. The purpose of this project is to investigate a novel design for light water reactor fuel which will extend fuel performance limits and improve reactor safety even further than is currently achieved. This project is investigating liquid metal bonding of LWR fuel in order to radically decrease fuel centerline temperatures which has major performance and safety benefits. The project will verify the compatibility of the liquid metal bond with both the fuel pellets and cladding material, verify the performance enhancement features of the new design over the fuel lifetime, and verify the economic fabricability of the concept and will show how this concept will benefit the LWR nuclear industry

  10. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    International Nuclear Information System (INIS)

    Winter, Dominik

    2014-01-01

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use of two fuel concepts based on a mixed oxide (MOX) fuel and an innovative thorium-plutonium (ThPu) fuel is investigated by a developed simulation model encompassing thermal hydraulics, neutronics, and fuel burnup. The main feature of these fuel concepts is the axially varying enrichment in plutonium which is, in this work, recycled from spent nuclear fuel and shows a high fission fraction of the absorption cross section for fast incident neutron energies. The potential of balancing the overall fuel utilization by an increase of the fission rate in the upper part of the active height with a combination of the harder spectrum and the higher fission fraction of the absorption cross section in the BWR core is studied. The three particular calculational models for thermal hydraulics, neutronics, and fuel burnup provide results at fuel assembly and/or at core level. In the former case, the main focus lies on the thermal hydraulics analysis, fuel burnup, and activity evolution after unloading from the core and, in the latter case, special attention is paid to reactivity safety coefficients (feedback effects) and the optimization of the operational behavior. At both levels (assembly and core), the isotopic buildup and depletion rates as a function of the active height are analyzed. In addition, a comparison between the use of conventional fuel types with homogeneous enrichments and the use of the innovative fuel types is made. In the framework of the simulations, the ThPu and the MOX

  11. Evaluation of Standard Concepts Design of Library Interior Physical Environment

    Directory of Open Access Journals (Sweden)

    Debri Harindya Putri

    2018-01-01

    Full Text Available Currently the function of a room is not only used as a shelter, the function of the room itself to be increased as a refreshing or relaxation area for users to follow the development of creativity and technology in the field of design. The comfortable factor becomes the main factor that indicates a successful process of creating a space. No exception library. The nature of library seemed stiff because of its function as a place to read, now can be developed and made into more dynamic with the special design concepts or color patterns used. Libraries can be created a special concept that suits the characteristics of the users themselves. Most users of the library, especially in college libraries are teenagers. Naturally, teenagers like to gather with their friends and we have to facilitate this activity in our library design concept. In addition we can also determine the needs of users through research by questionnaire method. The answers of users can be mapped and drawn conclusions. To explore the research, the author reviewed some literature about library interior design and observed the library of Ma Chung University as a case study. The combined results of the method can be concluded and the discovery of ideal standards of physical environment. So, the library can be made as a comfortable reading environment so as to increased interest in reading behavior and the frequent visits of students in the library

  12. Comparison of Design Concepts for SFR under Development

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Namduk; Choi, Yongwon; Bae, Moohoon; Shin, Andong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    The goal of ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) with a capacity of 600 MWe is to study the technical demonstration that can be scaled up to commercial reactor. It was expected that the success of ASTRID project could eventually lead to operation of industrial reactor around 2040. On 2012, ASTRID designer has submitted the DOrS (Dossier d’Orientations de Sûreté, Safety Orientation Document) for ASTRID to IRSN and IRSN has issued a report after reviewing the DOrS. The report DOrS itself is not available publicly, intellectual property might be the reason, but the review document of IRSN is open to public, so we can understand the basic concept of ASTRID by IRSN report. The DOrS of ASTRID and the TTR for PGSFR have not the same format and also the same purpose, so it is not easy to compare the two design concepts directly. But, still, we think the concepts could be compared in a very general way. Thus, in this paper we have presented the very short comparison results of the two SFR design. Our opinion after first reviewing the TTR is that the PGSFR needs to be designed in a more systematic way. The requirements are coming basically from the previous document used for SMART licensing and do not show prototype reactor specific characters.

  13. Development of environmentally advanced hydropower turbine system design concepts

    International Nuclear Information System (INIS)

    Franke, G.F.; Webb, D.R.; Fisher, R.K. Jr.

    1997-08-01

    A team worked together on the development of environmentally advanced hydro turbine design concepts to reduce hydropower''s impact on the environment, and to improve the understanding of the technical and environmental issues involved, in particular, with fish survival as a result of their passage through hydro power sites. This approach brought together a turbine design and manufacturing company, biologists, a utility, a consulting engineering firm and a university research facility, in order to benefit from the synergy of diverse disciplines. Through a combination of advanced technology and engineering analyses, innovative design concepts adaptable to both new and existing hydro facilities were developed and are presented. The project was divided into 4 tasks. Task 1 investigated a broad range of environmental issues and how the issues differed throughout the country. Task 2 addressed fish physiology and turbine physics. Task 3 investigated individual design elements needed for the refinement of the three concept families defined in Task 1. Advanced numerical tools for flow simulation in turbines are used to quantify characteristics of flow and pressure fields within turbine water passageways. The issues associated with dissolved oxygen enhancement using turbine aeration are presented. The state of the art and recent advancements of this technology are reviewed. Key elements for applying turbine aeration to improve aquatic habitat are discussed and a review of the procedures for testing of aerating turbines is presented. In Task 4, the results of the Tasks were assembled into three families of design concepts to address the most significant issues defined in Task 1. The results of the work conclude that significant improvements in fish passage survival are achievable

  14. Development of environmentally advanced hydropower turbine system design concepts

    Energy Technology Data Exchange (ETDEWEB)

    Franke, G.F.; Webb, D.R.; Fisher, R.K. Jr. [Voith Hydro, Inc. (United States)] [and others

    1997-08-01

    A team worked together on the development of environmentally advanced hydro turbine design concepts to reduce hydropower`s impact on the environment, and to improve the understanding of the technical and environmental issues involved, in particular, with fish survival as a result of their passage through hydro power sites. This approach brought together a turbine design and manufacturing company, biologists, a utility, a consulting engineering firm and a university research facility, in order to benefit from the synergy of diverse disciplines. Through a combination of advanced technology and engineering analyses, innovative design concepts adaptable to both new and existing hydro facilities were developed and are presented. The project was divided into 4 tasks. Task 1 investigated a broad range of environmental issues and how the issues differed throughout the country. Task 2 addressed fish physiology and turbine physics. Task 3 investigated individual design elements needed for the refinement of the three concept families defined in Task 1. Advanced numerical tools for flow simulation in turbines are used to quantify characteristics of flow and pressure fields within turbine water passageways. The issues associated with dissolved oxygen enhancement using turbine aeration are presented. The state of the art and recent advancements of this technology are reviewed. Key elements for applying turbine aeration to improve aquatic habitat are discussed and a review of the procedures for testing of aerating turbines is presented. In Task 4, the results of the Tasks were assembled into three families of design concepts to address the most significant issues defined in Task 1. The results of the work conclude that significant improvements in fish passage survival are achievable.

  15. Major aspects of the design of a first mirror for the ITER core CXRS diagnostics

    International Nuclear Information System (INIS)

    Krasikov, Yury; Panin, Anatoly; Biel, Wolfgang; Krimmer, Andreas; Litnovsky, Andrey; Mertens, Philippe; Neubauer, Olaf; Schrader, Michael

    2015-01-01

    Highlights: • Availability, technological issues, and changes in the ScMo structure to be solved in future. • Developed passively cooled mirror is a workable, flexible, scalable and robust concept. • The generic upper port plug is to be considerable customized. - Abstract: The ITER core charge exchange recombination spectroscopy diagnostics (cCXRS) occupies the vacuum vessel upper port #3 and includes, in its generic version, the following in-vessel components: an optical mirror system, a shutter, the diagnostic first wall and the neutron shielding block. The most vulnerable diagnostic mirror is obviously the first one (M1) directly observing the plasma. The M1 reference option is made of a single crystalline molybdenum (ScMo). The paper indicates major aspects influencing the first mirror design and identifies the most reasonable and reliable concept for cCXRS M1 at present. The applicability of the option presented is determined by many reasons, and especially, by the ITER generic upper port plug and its customization flexibility. The largest dimension of the mirror polished face is ∼300 mm. Such large ScMo workpieces are currently not available on the market. The mirror should be designed as an assembly of several ScMo pieces joined together. The M1 design is supported by multifield thermal, electromagnetic and structural analyses. The performed study confirms the feasibility of the proposed solutions. At the same time, the paper indicates numerous technological issues of the M1 unit to be solved in future.

  16. Major aspects of the design of a first mirror for the ITER core CXRS diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Krasikov, Yury, E-mail: y.krasikov@fz-juelich.de; Panin, Anatoly; Biel, Wolfgang; Krimmer, Andreas; Litnovsky, Andrey; Mertens, Philippe; Neubauer, Olaf; Schrader, Michael

    2015-10-15

    Highlights: • Availability, technological issues, and changes in the ScMo structure to be solved in future. • Developed passively cooled mirror is a workable, flexible, scalable and robust concept. • The generic upper port plug is to be considerable customized. - Abstract: The ITER core charge exchange recombination spectroscopy diagnostics (cCXRS) occupies the vacuum vessel upper port #3 and includes, in its generic version, the following in-vessel components: an optical mirror system, a shutter, the diagnostic first wall and the neutron shielding block. The most vulnerable diagnostic mirror is obviously the first one (M1) directly observing the plasma. The M1 reference option is made of a single crystalline molybdenum (ScMo). The paper indicates major aspects influencing the first mirror design and identifies the most reasonable and reliable concept for cCXRS M1 at present. The applicability of the option presented is determined by many reasons, and especially, by the ITER generic upper port plug and its customization flexibility. The largest dimension of the mirror polished face is ∼300 mm. Such large ScMo workpieces are currently not available on the market. The mirror should be designed as an assembly of several ScMo pieces joined together. The M1 design is supported by multifield thermal, electromagnetic and structural analyses. The performed study confirms the feasibility of the proposed solutions. At the same time, the paper indicates numerous technological issues of the M1 unit to be solved in future.

  17. A breed and burn reshuffling scheme for an Astrid-like reactor concept design

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois L, J. L., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2016-09-15

    The greenhouse emissions has a serious impact on environmental terms, reason why energy supply needs to be provided by low carbon emission technologies. Nuclear power is and environmentally friendly energy with future concepts and designs that ensure safety, and in this paper is taken as a solution to be considered in the energy supply mix. This study presents an extension on the operation of the Advanced Sodium Technological Reactor for Industrial demonstration (Astrid) nuclear reactor through a breed and burn operation to extend the operation of the reactor. The Astrid nuclear reactor is a fourth generation sodium-cooled fast reactor of 1500 MWt h, and it considers an innovative design: the low void effect core (Cfv: Coeur a Faible effet de Vidange sodium) due to the reactor configuration and the radial and axial position of the fuel subassemblies. Previous research in the Astrid-like cores aimed the model validation of a conventional oxide-fueled core and its comparison with a proposed metallic-fueled core. Taking into account the amount of fissile material (mainly 239-Pu) after the first cycle, a reshuffling scheme was suggested, which consists in changing strategically the position of the nuclear fuel assemblies when the reactivity drops near to the critical state of the reactor. Two different reshuffling schemes were simulated in every developed model, having operation extensions of 805 days and 1775 days for the oxide and metallic designs respectively. The implementation of the reshuffling schemes in the developed models enhanced the fuel utilization and could save up to 2.20 and 5.96 tons of plutonium for oxide and metallic designs respectively, which has an economic impact. The breeding of 239-Pu achieved in the fertile zone of the metallic design reached half of the initial concentration of the 239-Pu in the fissile zone and for the oxide design, the breeding reached one third of the initial concentration of the 239-Pu in the fissile zone. (Author)

  18. A breed and burn reshuffling scheme for an Astrid-like reactor concept design

    International Nuclear Information System (INIS)

    Garcia C, E. Y.; Francois L, J. L.

    2016-09-01

    The greenhouse emissions has a serious impact on environmental terms, reason why energy supply needs to be provided by low carbon emission technologies. Nuclear power is and environmentally friendly energy with future concepts and designs that ensure safety, and in this paper is taken as a solution to be considered in the energy supply mix. This study presents an extension on the operation of the Advanced Sodium Technological Reactor for Industrial demonstration (Astrid) nuclear reactor through a breed and burn operation to extend the operation of the reactor. The Astrid nuclear reactor is a fourth generation sodium-cooled fast reactor of 1500 MWt h, and it considers an innovative design: the low void effect core (Cfv: Coeur a Faible effet de Vidange sodium) due to the reactor configuration and the radial and axial position of the fuel subassemblies. Previous research in the Astrid-like cores aimed the model validation of a conventional oxide-fueled core and its comparison with a proposed metallic-fueled core. Taking into account the amount of fissile material (mainly 239-Pu) after the first cycle, a reshuffling scheme was suggested, which consists in changing strategically the position of the nuclear fuel assemblies when the reactivity drops near to the critical state of the reactor. Two different reshuffling schemes were simulated in every developed model, having operation extensions of 805 days and 1775 days for the oxide and metallic designs respectively. The implementation of the reshuffling schemes in the developed models enhanced the fuel utilization and could save up to 2.20 and 5.96 tons of plutonium for oxide and metallic designs respectively, which has an economic impact. The breeding of 239-Pu achieved in the fertile zone of the metallic design reached half of the initial concentration of the 239-Pu in the fissile zone and for the oxide design, the breeding reached one third of the initial concentration of the 239-Pu in the fissile zone. (Author)

  19. Design and Analysis of a Stiffened Composite Structure Repair Concept

    Science.gov (United States)

    Przekop, Adam

    2011-01-01

    A design and analysis of a repair concept applicable to a stiffened thin-skin composite panel based on the Pultruded Rod Stitched Efficient Unitized Structure is presented. Since the repair concept is a bolted repair using metal components, it can easily be applied in the operational environment. Initial analyses are aimed at validating the finite element modeling approach by comparing with available test data. Once confidence in the analysis approach is established several repair configurations are explored and the most efficient one presented. Repairs involving damage to the top of the stiffener alone are considered in addition to repairs involving a damaged stiffener, flange and underlying skin. High fidelity finite element modeling techniques such as mesh-independent definition of compliant fasteners, elastic-plastic metallic material properties and geometrically nonlinear analysis are utilized in the effort. The results of the analysis are presented and factors influencing the design are assessed and discussed.

  20. Automatic Voltage Control (AVC) of Danish Transmission System - Concept design

    DEFF Research Database (Denmark)

    Qin, Nan; Abildgaard, Hans; Lund, P.

    2014-01-01

    For more than 20 years it has been a consistent plan by all Danish governments to turn the Danish power production away from fossil fuels towards renewable energy. The result today is that 37% of the total Danish power consumption was covered by mainly wind energy in 2013 aiming at 50% by 2020......, objectives, constraints, algorithms for optimal power flow and some special functions in particular systems, which inspires the concept design of a Danish AVC system to address the future challenges of voltage control. In the concept, the Danish AVC design is based on a centralized control scheme. All...... the substation loses the telecommunications to the control center. RPCs will be integrated to the AVC system as normative regulators in the later stage. Distributed generation units can be organized as virtual power plants and participate in voltage control at transmission level. Energinet.dk as the Danish TSO...

  1. Design, construction and operating experience of demonstration LMFBRs. The application of core and fuel performance experience in British reactors to commercial fast reactor design

    International Nuclear Information System (INIS)

    Bagley, K.Q.

    1978-01-01

    The Prototype Fast Reactor (PFR) sub-assembly design is described with particular emphasis on the choice of factors that are important in determining satisfactory performance. Reasons for the adoption of specific clad and fuel design details are given in their historical context, and irradiation experience - mostly from the Dounreay Fast Reactor (DFR) - in support of the choices is described. The implications of factors that are now better understood than when the PFR fuel was designed, notably neutron-induced void swelling and irradiation creep, are then considered. It is shown that the 'free-standing' core design used in PFR, in which the sub-assembly is unsupported above the level of the lower axial breeder, relies on the availability of low-swelling, preferably irradiation-creep-resistant alloys as sub-assembly structural materials in order to achieve the prescribed burn-up target. The advantages of a 'restrained core', which makes use of irradiation creep to redress the effects of material swelling, are noted briefly, and the application of this concept to the Commercial Demonstration Fast Reactor (CDFR) core design is described. Probable future trends in pin and sub-assembly design are reviewed and the scope of associated irradiation testing programmes defined. Arrangements for monitoring and evaluating fuel performance, both in reactor and post-irradiation, are outlined and the provisions for endorsement of CDFR pin, sub-assembly and core design details in PFR are indicated. (author)

  2. Licensing aspects in the verification of the SNR 300 design concept against hypothetical accidents

    International Nuclear Information System (INIS)

    Kugler, E.; Wiesner, S.

    1976-01-01

    The German prototype of a fast breeder reactor, the SNR 300, is being built near Kalkar on the Lower Rhine. It is a loop-type fast sodium-cooled reactor, designed and constructed by Interatom, Bensberg. Experiences gained from the first phase of construction are described. The report is restricted to the aspects of the SNR 300 design against a core disruptive accident (CDA) and its consequences and to the difficulties having arisen in the verification of the design concept so far. Some examples of the detailed design are described and discussed from the licensing authority's point of view showing that the difficulties have been typical for a prototype reactor subjected to a regular licensing procedure

  3. IT-tool Concept for Design and Intelligent Motion Control

    DEFF Research Database (Denmark)

    Conrad, Finn; Hansen, Poul Erik; Sørensen, Torben

    2000-01-01

    The paper presents results obtained from a Danish mechatronic research program focusing on intelligent motion control as well as results from the Esprit project SWING on IT-tools for rapid prototyping of fluid power components and systems. A mechatronic test facility with digital controllers for ....... Furthermore, a developed IT-tool concept for controller and system design utilising the ISO 10303 STEP Standard is proposed....

  4. Fuel transfer cask concept design for reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Ahmad Nabil Ab Rahim; Phongsakorn Prak; Tonny Lanyau; Mohd Fazli Zakaria

    2010-01-01

    Reactor Triga PUSPATI (RTP) has been operated since 1982 till now. For such long period, the organization feels the need to upgrade the power from 1 MW to 3 MW which involved changing new fuels. Spent fuels will be stored in a Spent Fuel Pool. The process of transferring spent fuels into Spent Fuels Pool required a fuel transfer cask. This paper discussed the design concept for the fuel transfer cast which is essential equipment for reactor upgrading mission. (author)

  5. OSU TOMF Program Site Selection and Preliminary Concept Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Spadling, Steve [Oklahoma State Univ., Stillwater, OK (United States)

    2012-05-10

    The purpose of this report is to confirm the programmatic requirements for the new facilities, identify the most appropriate project site, and develop preliminary site and building concepts that successfully address the overall project goals and site issues. These new facilities will be designed to accommodate the staff, drivers and maintenance requirements for the future mixed fleet of passenger vehicles, Transit Style Buses and School Buses.

  6. Safeguards by design - The early consideration of safeguards concepts

    International Nuclear Information System (INIS)

    Killeen, T.; Moran, B.; Pujol, E.

    2009-01-01

    Full-text: The IAEA Department of Safeguards is in the process of formalizing its approach to long-range strategic planning. As a result of this activity new endeavours are being identified. One of these endeavours is to develop a concept known as Safeguards by Design. Safeguarding nuclear material and facilities can be made more effective and cost efficient by improving the safeguardability of the system. By taking into account design features that facilitate the implementation of international safeguards early in the design phase, a concept known as safeguards by design, the proliferation resistance of the system can be improved. This improvement process requires an understanding by designers and operators of safeguards and its underlying principles. To advance the safeguards by design approach, the IAEA determined that there is a need to develop written guidance. This guidance would help the major stakeholders - the designers, operators, owners, and regulatory bodies - to better understand how a facility could be designed, built and operated in such a way that effective safeguards could be implemented at reduced cost and with minimal burden to facility operations. By enlisting the cooperation of Member States through the support programme structure, the IAEA is working to first develop a document that describes the basic principles of safeguards, and the fundamental design features and measures that facilitate the implementation of international safeguards. Facility-specific guidance will then be developed utilizing the resources, expertise and experience of the IAEA and its Member States. This paper will review the foundation for the development of this task, describe the progress that has been made and outline the path forward. (author)

  7. Optimal burnable poison utilization in PWR core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.

    1986-01-01

    A method was developed for determining the optimal distribution and depletion of burnable poisons in a Pressurized Water Reactor core. The well-known Haling depletion technique is used to achieve the end-of-cycle core state where the fuel assembly arrangement is configured in the absence of all control poison. The soluble and burnable poison required to control the core reactivity and power distribution are solved for as unknown variables while step depleting the cycle in reverse with a target power distribution. The method was implemented in the NRC approved licensing code SIMULATE

  8. Design of the core support and restraint structures for FFTF and CRBRP

    International Nuclear Information System (INIS)

    Sutton, H.G.; Rylatt, J.A.

    1977-12-01

    This paper presents and compares the design and fabrication of the FFTF and CRBRP reactor structures which support and restrain the reactor core assemblies. The fabrication of the core support structure (CSS) for the FFTF reactor was completed October 1972 and this paper discusses how the fabrication problems encountered with the FFTF were avoided in the subsequent design of the CRBR CSS. The radial core restraint structure of the FFTF was designed and fabricated such that an active system could replace the present passive system which is segmented and relies on the CSS core barrel for total structure integrity to maintain core geometry. The CRBR core restraint structure is designed for passive restraint only, and this paper discusses how the combined strengths of the restraint structure former rings and the CSS core barrel are utilized to maintain core geometry. Whereas the CSS for the FFTF interfaces directly with the reactor core assemblies, the CRBR CSS does not. A comparison is made on how intermediate structures in CRBR (inlet modules) provide the necessary design interfaces for supporting and providing flow distribution to the reactor core assemblies. A discussion is given on how the CRBR CSS satisfied the design requirements of the Equipment Specification, including thermal transient, dynamic and seismic loadings, and results of flow distribution testing that supported the CRBR design effort. The approach taken to simplify fabrication of the CRBR components, and a novel 20 inch deep narrow gap weld joint in the CSS are described

  9. Generic repository design concepts and thermal analysis (FY11)

    International Nuclear Information System (INIS)

    Howard, Robert; Dupont, Mark; Blink, James A.; Fratoni, Massimiliano; Greenberg, Harris; Carter, Joe; Hardin, Ernest L.; Sutton, Mark A.

    2011-01-01

    Reference concepts for geologic disposal of used nuclear fuel and high-level radioactive waste in the U.S. are developed, including geologic settings and engineered barriers. Repository thermal analysis is demonstrated for a range of waste types from projected future, advanced nuclear fuel cycles. The results show significant differences among geologic media considered (clay/shale, crystalline rock, salt), and also that waste package size and waste loading must be limited to meet targeted maximum temperature values. In this study, the UFD R and D Campaign has developed a set of reference geologic disposal concepts for a range of waste types that could potentially be generated in advanced nuclear FCs. A disposal concept consists of three components: waste inventory, geologic setting, and concept of operations. Mature repository concepts have been developed in other countries for disposal of spent LWR fuel and HLW from reprocessing UNF, and these serve as starting points for developing this set. Additional design details and EBS concepts will be considered as the reference disposal concepts evolve. The waste inventory considered in this study includes: (1) direct disposal of SNF from the LWR fleet, including Gen III+ advanced LWRs being developed through the Nuclear Power 2010 Program, operating in a once-through cycle; (2) waste generated from reprocessing of LWR UOX UNF to recover U and Pu, and subsequent direct disposal of used Pu-MOX fuel (also used in LWRs) in a modified-open cycle; and (3) waste generated by continuous recycling of metal fuel from fast reactors operating in a TRU burner configuration, with additional TRU material input supplied from reprocessing of LWR UOX fuel. The geologic setting provides the natural barriers, and establishes the boundary conditions for performance of engineered barriers. The composition and physical properties of the host medium dictate design and construction approaches, and determine hydrologic and thermal responses of

  10. Generic repository design concepts and thermal analysis (FY11).

    Energy Technology Data Exchange (ETDEWEB)

    Howard, Robert (Oak Ridge National Laboratory, Oak Ridge, TN); Dupont, Mark (Savannah River Nuclear Solutions, Aiken, SC); Blink, James A. (Lawrence Livermore National Laboratory, Livermore, CA); Fratoni, Massimiliano (Lawrence Livermore National Laboratory, Livermore, CA); Greenberg, Harris (Lawrence Livermore National Laboratory, Livermore, CA); Carter, Joe (Savannah River Nuclear Solutions, Aiken, SC); Hardin, Ernest L.; Sutton, Mark A. (Lawrence Livermore National Laboratory, Livermore, CA)

    2011-08-01

    Reference concepts for geologic disposal of used nuclear fuel and high-level radioactive waste in the U.S. are developed, including geologic settings and engineered barriers. Repository thermal analysis is demonstrated for a range of waste types from projected future, advanced nuclear fuel cycles. The results show significant differences among geologic media considered (clay/shale, crystalline rock, salt), and also that waste package size and waste loading must be limited to meet targeted maximum temperature values. In this study, the UFD R&D Campaign has developed a set of reference geologic disposal concepts for a range of waste types that could potentially be generated in advanced nuclear FCs. A disposal concept consists of three components: waste inventory, geologic setting, and concept of operations. Mature repository concepts have been developed in other countries for disposal of spent LWR fuel and HLW from reprocessing UNF, and these serve as starting points for developing this set. Additional design details and EBS concepts will be considered as the reference disposal concepts evolve. The waste inventory considered in this study includes: (1) direct disposal of SNF from the LWR fleet, including Gen III+ advanced LWRs being developed through the Nuclear Power 2010 Program, operating in a once-through cycle; (2) waste generated from reprocessing of LWR UOX UNF to recover U and Pu, and subsequent direct disposal of used Pu-MOX fuel (also used in LWRs) in a modified-open cycle; and (3) waste generated by continuous recycling of metal fuel from fast reactors operating in a TRU burner configuration, with additional TRU material input supplied from reprocessing of LWR UOX fuel. The geologic setting provides the natural barriers, and establishes the boundary conditions for performance of engineered barriers. The composition and physical properties of the host medium dictate design and construction approaches, and determine hydrologic and thermal responses of the

  11. Fuel management strategy for the compact core design of RSG GAS (MPR-30)

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, T.M.; Liem, P.H.; Tukiran, S. [National Nuclear Energy Agency (Batan), PUSPIPTEK-Serpong Tangerang (Indonesia)

    2000-07-01

    The rearrangement of the core configuration of the RSG GAS reactor to obtain a compact core is in progress. A fuel management strategy is proposed for the equilibrium compact core of this reactor by reducing the number of in-core irradiation positions. The reduced irradiation positions are based on the activities during 12 years operation. The obtained compact core gives significant extension of the operation cycle length so that the reactor availability and utilization can be enhanced. The equilibrium compact silicide core obtained met the imposed design constraints and safety requirements. (author)

  12. Fuel management strategy for the compact core design of RSG GAS (MPR-30)

    International Nuclear Information System (INIS)

    Sembiring, T.M.; Liem, P.H.; Tukiran, S.

    2000-01-01

    The rearrangement of the core configuration of the RSG GAS reactor to obtain a compact core is in progress. A fuel management strategy is proposed for the equilibrium compact core of this reactor by reducing the number of in-core irradiation positions. The reduced irradiation positions are based on the activities during 12 years operation. The obtained compact core gives significant extension of the operation cycle length so that the reactor availability and utilization can be enhanced. The equilibrium compact silicide core obtained met the imposed design constraints and safety requirements. (author)

  13. Automating expert role to determine design concept in Kansei Engineering

    Science.gov (United States)

    Lokman, Anitawati Mohd; Haron, Mohammad Bakri Che; Abidin, Siti Zaleha Zainal; Khalid, Noor Elaiza Abd

    2016-02-01

    Affect has become imperative in product quality. In affective design field, Kansei Engineering (KE) has been recognized as a technology that enables discovery of consumer's emotion and formulation of guide to design products that win consumers in the competitive market. Albeit powerful technology, there is no rule of thumb in its analysis and interpretation process. KE expertise is required to determine sets of related Kansei and the significant concept of emotion. Many research endeavors become handicapped with the limited number of available and accessible KE experts. This work is performed to simulate the role of experts with the use of Natphoric algorithm thus providing sound solution to the complexity and flexibility in KE. The algorithm is designed to learn the process by implementing training datasets taken from previous KE research works. A framework for automated KE is then designed to realize the development of automated KE system. A comparative analysis is performed to determine feasibility of the developed prototype to automate the process. The result shows that the significant Kansei is determined by manual KE implementation and the automated process is highly similar. KE research advocates will benefit this system to automatically determine significant design concepts.

  14. Uncertainty Evaluation of Reactivity Coefficients for a large advanced SFR Core Design

    International Nuclear Information System (INIS)

    Khamakhem, Wassim; Rimpault, Gerald

    2008-01-01

    Sodium Cooled Fast Reactors are currently being reshaped in order to meet Generation IV goals on economics, safety and reliability, sustainability and proliferation resistance. Recent studies have led to large SFR cores for a 3600 MWth power plants, cores which exhibit interesting features. The designs have had to balance between competing aspects such as sustainability and safety characteristics. Sustainability in neutronic terms is translated into positive breeding gain and safety into rather low Na void reactivity effects. The studies have been done on two SFR concepts using oxide and carbide fuels. The use of the sensitivity theory in the ERANOS determinist code system has been used. Calculations have been performed with different sodium evaluations: JEF2.2, ERALIB-1 and the most recent JEFF3.1 and ENDF/B-VII in order to make a broad comparison. Values for the Na void reactivity effect exhibit differences as large as 14% when using the different sodium libraries. Uncertainties due to nuclear data on the reactivity coefficients were performed with BOLNA variances-covariances data, the Na Void Effect uncertainties are near to 12% at 1σ. Since, the uncertainties are far beyond the target accuracy for a design achieving high performance, two directions are envisaged: the first one is to perform new differential measurements or in a second attempt use integral experiments to improve effectively the nuclear data set and its uncertainties such as performed in the past with ERALIB1. (authors)

  15. Application of core structural design guidelines in conceptual fuel pin design

    International Nuclear Information System (INIS)

    Patel, M.R.; Stephen, J.D.

    1979-01-01

    The paper describes an application of the Draft RDT Standards F9-7, -8, and -9 to conceptual design of Fast Breeder Reactor (FBR) fuel pins. The Standards are being developed to provide guidelines for structural analysis and design of the FBR core components which have limited ductility at high fluences and are not addressed by the prevalent codes. The development is guided by a national working group sponsored by the Division of Reactor Researcch and Technology of the Department of Energy. The development program summarized in the paper includes establishment of design margins consistent with the test data and component performance requirements, and application of the design rules in various design activities. The application program insures that the quantities required for proper application of the design rules are available from the analysis methods and test data, and that the use of the same design rules in different analysis tools used at different stages of a component design producees consistent results. This is illustrated in the paper by application of the design rules in the analysis methods developed for conceptual and more detailed designs of an FBR fuel pin

  16. Application of neural network to multi-dimensional design window search in reactor core design

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki

    1999-01-01

    In the reactor core design, many parametric survey calculations should be carried out to decide an optimal set of basic design parameter values. They consume a large amount of computation time and labor in the conventional way. To support design work, we investigate a procedure to search efficiently a design window, which is defined as feasible design parameter ranges satisfying design criteria and requirements, in a multi-dimensional space composed of several basic design parameters. The present method is applied to the neutronics and thermal hydraulics fields. The principle of the present method is to construct the multilayer neural network to simulate quickly a response of an analysis code through a training process, and to reduce computation time using the neural network without parametric study using analysis codes. To verify the applicability of the present method to the neutronics and the thermal hydraulics design, we have applied it to high conversion water reactors and examined effects of the structure of the neural network and the number of teaching patterns on the accuracy of the design window estimated by the neural network. From the results of the applications, a guideline to apply the present method is proposed and the present method can predict an appropriate design window in a reasonable computation time by following the guideline. (author)

  17. Feasibility study on nuclear core design for soluble boron free small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Hah, Chang Joo; Ju, Cho Sung [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  18. Pulsed air-core deflector-magnet design parameters

    International Nuclear Information System (INIS)

    Jason, A.J.; Cooper, R.K.; Liebman, A.D.; Blind, B.; Koelle, A.R.

    1983-01-01

    The response of air-core magnets to pulsed excitation is dependent on the pulse frequency spectrum because of fields produced by induced currents in the magnet structure. We discuss this phenomenon quantitatively in terms of magnet performance optimization

  19. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Boer, Brian

    2007-01-01

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  20. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K.

    2012-09-15

    plenum and boron layer, varying the core height-to-diameter (H/D) ratio, introducing moderator pins into the fuel assemblies, and modifying the initially loaded plutonium content. The sensitivity of the principal safety and performance parameters, viz. void reactivity, Doppler constant, nominal reactivity and breeding gain, has been evaluated with respect to each of the options. Finally, two synthesis core concepts, representing different selected combinations of the optimization options, have been proposed. The third and last phase has been to test the proposed optimized designs in terms of the dynamic core response to a representative ULOF accident, in which sodium boiling occurs. Such a hypothetical transient is of prime importance for SFR safety demonstration, since it may lead to a severe accident situation, where both cladding and fuel melt. As compared to the response of the reference ESFR core, certain improvements are indeed achieved with the static neutronics optimization carried out. However, these are found, in themselves, to be insufficient as regards the prevention of cladding and fuel melting. Thermal-hydraulic optimization has thus been considered necessary to: 1) prevent sodium flow blockage in the fuel channels and 2) avoid boiling instabilities caused by the vaporization/condensation process in the upper sodium plenum. Following implementation of appropriate thermal-hydraulics related design changes, one arrives at a final configuration of the SFR core in which, for the selected accident scenario, a new ‘steady state’ involving stable sodium boiling is shown to be achievable, with melting of neither cladding nor fuel. Such satisfactory behavior has been confirmed not only for the beginning-of-life core state, but also for the equilibrium closed fuel cycle. (author)

  1. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Sun, K.

    2012-09-01

    and boron layer, varying the core height-to-diameter (H/D) ratio, introducing moderator pins into the fuel assemblies, and modifying the initially loaded plutonium content. The sensitivity of the principal safety and performance parameters, viz. void reactivity, Doppler constant, nominal reactivity and breeding gain, has been evaluated with respect to each of the options. Finally, two synthesis core concepts, representing different selected combinations of the optimization options, have been proposed. The third and last phase has been to test the proposed optimized designs in terms of the dynamic core response to a representative ULOF accident, in which sodium boiling occurs. Such a hypothetical transient is of prime importance for SFR safety demonstration, since it may lead to a severe accident situation, where both cladding and fuel melt. As compared to the response of the reference ESFR core, certain improvements are indeed achieved with the static neutronics optimization carried out. However, these are found, in themselves, to be insufficient as regards the prevention of cladding and fuel melting. Thermal-hydraulic optimization has thus been considered necessary to: 1) prevent sodium flow blockage in the fuel channels and 2) avoid boiling instabilities caused by the vaporization/condensation process in the upper sodium plenum. Following implementation of appropriate thermal-hydraulics related design changes, one arrives at a final configuration of the SFR core in which, for the selected accident scenario, a new ‘steady state’ involving stable sodium boiling is shown to be achievable, with melting of neither cladding nor fuel. Such satisfactory behavior has been confirmed not only for the beginning-of-life core state, but also for the equilibrium closed fuel cycle. (author)

  2. MO-E-18C-04: Advanced Computer Simulation and Visualization Tools for Enhanced Understanding of Core Medical Physics Concepts

    International Nuclear Information System (INIS)

    Naqvi, S

    2014-01-01

    Purpose: Most medical physics programs emphasize proficiency in routine clinical calculations and QA. The formulaic aspect of these calculations and prescriptive nature of measurement protocols obviate the need to frequently apply basic physical principles, which, therefore, gradually decay away from memory. E.g. few students appreciate the role of electron transport in photon dose, making it difficult to understand key concepts such as dose buildup, electronic disequilibrium effects and Bragg-Gray theory. These conceptual deficiencies manifest when the physicist encounters a new system, requiring knowledge beyond routine activities. Methods: Two interactive computer simulation tools are developed to facilitate deeper learning of physical principles. One is a Monte Carlo code written with a strong educational aspect. The code can “label” regions and interactions to highlight specific aspects of the physics, e.g., certain regions can be designated as “starters” or “crossers,” and any interaction type can be turned on and off. Full 3D tracks with specific portions highlighted further enhance the visualization of radiation transport problems. The second code calculates and displays trajectories of a collection electrons under arbitrary space/time dependent Lorentz force using relativistic kinematics. Results: Using the Monte Carlo code, the student can interactively study photon and electron transport through visualization of dose components, particle tracks, and interaction types. The code can, for instance, be used to study kerma-dose relationship, explore electronic disequilibrium near interfaces, or visualize kernels by using interaction forcing. The electromagnetic simulator enables the student to explore accelerating mechanisms and particle optics in devices such as cyclotrons and linacs. Conclusion: The proposed tools are designed to enhance understanding of abstract concepts by highlighting various aspects of the physics. The simulations serve as

  3. MO-E-18C-04: Advanced Computer Simulation and Visualization Tools for Enhanced Understanding of Core Medical Physics Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Naqvi, S [Saint Agnes Cancer Institute, Department of Radiation Oncology, Baltimore, MD (United States)

    2014-06-15

    Purpose: Most medical physics programs emphasize proficiency in routine clinical calculations and QA. The formulaic aspect of these calculations and prescriptive nature of measurement protocols obviate the need to frequently apply basic physical principles, which, therefore, gradually decay away from memory. E.g. few students appreciate the role of electron transport in photon dose, making it difficult to understand key concepts such as dose buildup, electronic disequilibrium effects and Bragg-Gray theory. These conceptual deficiencies manifest when the physicist encounters a new system, requiring knowledge beyond routine activities. Methods: Two interactive computer simulation tools are developed to facilitate deeper learning of physical principles. One is a Monte Carlo code written with a strong educational aspect. The code can “label” regions and interactions to highlight specific aspects of the physics, e.g., certain regions can be designated as “starters” or “crossers,” and any interaction type can be turned on and off. Full 3D tracks with specific portions highlighted further enhance the visualization of radiation transport problems. The second code calculates and displays trajectories of a collection electrons under arbitrary space/time dependent Lorentz force using relativistic kinematics. Results: Using the Monte Carlo code, the student can interactively study photon and electron transport through visualization of dose components, particle tracks, and interaction types. The code can, for instance, be used to study kerma-dose relationship, explore electronic disequilibrium near interfaces, or visualize kernels by using interaction forcing. The electromagnetic simulator enables the student to explore accelerating mechanisms and particle optics in devices such as cyclotrons and linacs. Conclusion: The proposed tools are designed to enhance understanding of abstract concepts by highlighting various aspects of the physics. The simulations serve as

  4. A concept ideation framework for medical device design.

    Science.gov (United States)

    Hagedorn, Thomas J; Grosse, Ian R; Krishnamurty, Sundar

    2015-06-01

    Medical device design is a challenging process, often requiring collaboration between medical and engineering domain experts. This collaboration can be best institutionalized through systematic knowledge transfer between the two domains coupled with effective knowledge management throughout the design innovation process. Toward this goal, we present the development of a semantic framework for medical device design that unifies a large medical ontology with detailed engineering functional models along with the repository of design innovation information contained in the US Patent Database. As part of our development, existing medical, engineering, and patent document ontologies were modified and interlinked to create a comprehensive medical device innovation and design tool with appropriate properties and semantic relations to facilitate knowledge capture, enrich existing knowledge, and enable effective knowledge reuse for different scenarios. The result is a Concept Ideation Framework for Medical Device Design (CIFMeDD). Key features of the resulting framework include function-based searching and automated inter-domain reasoning to uniquely enable identification of functionally similar procedures, tools, and inventions from multiple domains based on simple semantic searches. The significance and usefulness of the resulting framework for aiding in conceptual design and innovation in the medical realm are explored via two case studies examining medical device design problems. Copyright © 2015 Elsevier Inc. All rights reserved.

  5. Design and analysis of advanced flight planning concepts

    Science.gov (United States)

    Sorensen, John A.

    1987-01-01

    The objectives of this continuing effort are to develop and evaluate new algorithms and advanced concepts for flight management and flight planning. This includes the minimization of fuel or direct operating costs, the integration of the airborne flight management and ground-based flight planning processes, and the enhancement of future traffic management systems design. Flight management (FMS) concepts are for on-board profile computation and steering of transport aircraft in the vertical plane between a city pair and along a given horizontal path. Flight planning (FPS) concepts are for the pre-flight ground based computation of the three-dimensional reference trajectory that connects the city pair and specifies the horizontal path, fuel load, and weather profiles for initializing the FMS. As part of these objectives, a new computer program called EFPLAN has been developed and utilized to study advanced flight planning concepts. EFPLAN represents an experimental version of an FPS. It has been developed to generate reference flight plans compatible as input to an FMS and to provide various options for flight planning research. This report describes EFPLAN and the associated research conducted in its development.

  6. Uncertainty reevaluation of T/H parameters of HANARO core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hark Rho; Park, Cheol; Kim, Heo Nil; Chae, Hee Taek

    1999-03-01

    HANARO core was designed by statistical thermal design method which was generally applied to power plant design. However, reevaluation of core thermal margin reflecting design changes as well as experiences through commissioning and operation is necessary for safe operation of reactor. For this objective, the revision of data for T/H design parameters and the reevaluation of their uncertainties were performed. (Author). 30 refs., 7 figs.

  7. Study on Reduced-Moderation Water Reactor (RMWR) core design. Joint research report (FY1998-1999)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-09-01

    The Reduce-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor aiming at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. Japan Atomic Energy Research Institute (JAERI) started a joint research program for conceptual design of RMWR core in collaboration with the Japan Atomic Power Company (JAPC) since 1998. The research area includes the RMWR core conceptual designs, development of analysis methods for rector physics and thermal-hydraulics to design the RMWR cores with higher accuracy and preparation of MOX critical experiment to confirm the feasibility from the reactor physics point of view. The present report describes the results of joint research program 'RMWR core design Phase 1' performed by JAERI and JAPC in FY 1998 and 1999. The results obtained from the joint research program are as follows: Conceptual design study on the RMWR core has been performed. A core concept with a conversion ratio more than about 1 is basically feasible to multiple recycling of plutonium. Investigating core characteristics at the equilibrium, some promising core concepts to satisfy above aims have been established. As for BWR-type concepts with negative void reactivity coefficients, three types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.1, (2) one feasible to attain operation cycle of about 2 years and burn-up of about 60 GWd/t with conversion ratio more than 1 or (3) one in simple design based on the ABWR assembly and without blanket attaining conversion ratio more than 1. And as for PWR-type concepts with negative void reactivity coefficients, two types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.05 by using heavy water as a coolant and (2) one feasible to attain conversion ratio about l by using light water. In the study of nuclear calculation method, a reactor analysis code

  8. Addressing Pre-Service Teachers' Understandings and Difficulties with Some Core Concepts in the Special Theory of Relativity

    Science.gov (United States)

    Selcuk, Gamze Sezgin

    2011-01-01

    The aim of this study is to investigate pre-service teachers' understanding of and difficulties with some core concepts in the special theory of relativity. The pre-service teachers (n = 185) from the Departments of Physics Education and Elementary Science Education at Dokuz Eylul University (in Turkey) participated. Both quantitative and…

  9. The Curriculum-Faculty-Reinforcement Alignment and Its Effect on Learning Retention of Core Marketing Concepts of Marketing Capstone Students

    Science.gov (United States)

    Raska, David; Keller, Eileen Weisenbach; Shaw, Doris

    2014-01-01

    Curriculum-Faculty-Reinforcement (CFR) alignment is an alignment between fundamental marketing concepts that are integral to the mastery of knowledge expected of our marketing graduates, their perceived importance by the faculty, and their level of reinforcement throughout core marketing courses required to obtain a marketing degree. This research…

  10. RASC-AL (Revolutionary Aerospace Systems Concepts-Academic Linkage): 2002 Advanced Concept Design Presentation

    Science.gov (United States)

    2002-01-01

    The Revolutionary Aerospace Systems Concepts-Academic Linkage (RASC-AL) is a program of the Lunar and Planetary Institute (LPI) in collaboration with the Universities Space Research Association's (USRA) ICASE institute through the NASA Langley Research Center. The RASC-AL key objectives are to develop relationships between universities and NASA that lead to opportunities for future NASA research and programs, and to develop aerospace systems concepts and technology requirements to enable future NASA missions. The program seeks to look decades into the future to explore new mission capabilities and discover what's possible. NASA seeks concepts and technologies that can make it possible to go anywhere, at anytime, safely, reliably, and affordably to accomplish strategic goals for science, exploration, and commercialization. University teams were invited to submit research topics from the following themes: Human and Robotic Space Exploration, Orbital Aggregation & Space Infrastructure Systems (OASIS), Zero-Emissions Aircraft, and Remote Sensing. RASC-AL is an outgrowth of the HEDS-UP (University Partners) Program sponsored by the LPI. HEDS-UP was a program of the Lunar and Planetary Institute designed to link universities with NASA's Human Exploration and Development of Space (HEDS) enterprise. The first RASC-AL Forum was held November 5-8, 2002, at the Hilton Cocoa Beach Oceanfront Hotel in Cocoa Beach, Florida. Representatives from 10 university teams presented student research design projects at this year's Forum. Each team contributed a written report and these reports are presented.

  11. Design Concepts for Cooled Ceramic Composite Turbine Vane

    Science.gov (United States)

    Boyle, Robert J.; Parikh, Ankur H.; Nagpal, VInod K.

    2015-01-01

    The objective of this work was to develop design concepts for a cooled ceramic vane to be used in the first stage of the High Pressure Turbine(HPT). To insure that the design concepts were relevant to the gas turbine industry needs, Honeywell International Inc. was subcontracted to provide technical guidance for this work. The work performed under this contract can be divided into three broad categories. The first was an analysis of the cycle benefits arising from the higher temperature capability of Ceramic Matrix Composite(CMC) compared with conventional metallic vane materials. The second category was a series of structural analyses for variations in the internal configuration of first stage vane for the High Pressure Turbine(HPT) of a CF6 class commercial airline engine. The third category was analysis for a radial cooled turbine vanes for use in turboshaft engine applications. The size, shape and internal configuration of the turboshaft engine vanes were selected to investigate a cooling concept appropriate to small CMC vanes.

  12. Methodology for thermal hydraulic conceptual design and performance analysis of KALIMER core

    International Nuclear Information System (INIS)

    Young-Gyun Kim; Won-Seok Kim; Young-Jin Kim; Chang-Kue Park

    2000-01-01

    This paper summarizes the methodology for thermal hydraulic conceptual design and performance analysis which is used for KALIMER core, especially the preliminary methodology for flow grouping and peak pin temperature calculation in detail. And the major technical results of the conceptual design for the KALIMER 98.03 core was shown and compared with those of KALIMER 97.07 design core. The KALIMER 98.03 design core is proved to be more optimized compared to the 97.07 design core. The number of flow groups are reduced from 16 to 11, and the equalized peak cladding midwall temperature from 654 deg. C to 628 deg. C. It was achieved from the nuclear and thermal hydraulic design optimization study, i.e. core power flattening and increase of radial blanket power fraction. Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in core thermal hydraulic analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each bundle, thus pin cladding damage accumulation and pin reliability. The flow grouping and the peak pin temperature calculation for the preliminary conceptual design is performed with the modules ORFCE-F60 and ORFCE-T60 respectively. The basic subchannel analysis will be performed with the SLTHEN code, and the detailed subchannel analysis will be done with the MATRA-LMR code which is under development for the K-Core system. This methodology was proved practical to KALIMER core thermal hydraulic design from the related benchmark calculation studies, and it is used to KALIMER core thermal hydraulic conceptual design. (author)

  13. Molten Core - Concrete interactions in nuclear accidents. Theory and design of an experimental facility

    International Nuclear Information System (INIS)

    Sevon, T.

    2005-11-01

    In a hypothetical severe accident in a nuclear power plant, the molten core of the reactor may flow onto the concrete floor of containment building. This would cause a molten core . concrete interaction (MCCI), in which the heat transfer from the hot melt to the concrete would cause melting of the concrete. In assessing the safety of nuclear reactors, it is important to know the consequences of such an interaction. As background to the subject, this publication includes a description of the core melt stabilization concept of the European Pressurized water Reactor (EPR), which is being built in Olkiluoto in Finland. The publication includes a description of the basic theory of the interaction and the process of spalling or cracking of concrete when it is heated rapidly. A literature survey and some calculations of the physical properties of concrete and corium. concrete mixtures at high temperatures have been conducted. In addition, an equation is derived for conservative calculation of the maximum possible concrete ablation depth. The publication also includes a literature survey of experimental research on the subject of the MCCI and discussion of the results and deficiencies of the experiments. The main result of this work is the general design of an experimental facility to examine the interaction of molten metals and concrete. The main objective of the experiments is to assess the probability of spalling, or cracking, of concrete under pouring of molten material. A program of five experiments has been designed, and pre-test calculations of the experiments have been conducted with MELCOR 1.8.5 accident analysis program and conservative analytic calculations. (orig.)

  14. Improving fuel utilization in SmAHTR with spectral shift control design: Proof of concept

    International Nuclear Information System (INIS)

    Kotlyar, D.; Lindley, B.A.; Mohamed, H.

    2017-01-01

    Highlights: • Improving the fuel utilization in a graphite moderated reactor by adopting the ‘spectral shift’ concept. • The feasibility of this concept was tested in the Small Advanced High-Temperature Reactor. • At BOL, the reactor is under-moderated, with excess neutrons being primarily breeding 239 Pu. • Graphite is continuously inserted thermalizing the neutron spectrum and increasing reactivity. • The extra 239 Pu bred during the cycle is then burned, allowing the cycle to be extended. - Abstract: This paper presents a spectral shift design based approach to improve the fuel utilization factor or alternatively to increase the cycle length in a graphite moderated reactor. The feasibility of this concept was tested in the Small Advanced High-Temperature Reactor (SmAHTR). This is a small sized Fluoride-salt-cooled high-temperature reactor (FHR) that uses tri-isotropic (TRISO)-coated particle fuels and graphite moderator materials. A major benefit of the TRISO particles is the ability to mitigate fission product release in the case of an accident. However, the fabrication costs associated with TRISO particles are expected to be significantly higher than the traditional UO 2 fuel. The preliminary studies presented in the paper are focused on extending the achievable irradiation period without increasing the value of the enrichment. In order to increase the discharge burnup, the design includes graphite structures that are initially removed from the core. This imposes a harder spectrum, which enhances the breeding of 239 Pu. Then, the graphite structures are gradually and continuously inserted into the core to sustain criticality. This procedure shifts the hard spectrum into a more thermal one and enables a more efficient utilization of 239 Pu. The preliminary results indicate that this design achieves considerably longer irradiation periods and hence lower fuel cycle costs than the reference design.

  15. Development of small, fast reactor core designs using lead-based coolant

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Hill, R. N.; Khalil, H. S.; Wade, D. C.

    1999-01-01

    A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations

  16. Design and analysis of three-layer-core optical fiber

    Science.gov (United States)

    Zheng, Siwen; Liu, Yazhuo; Chang, Guangjian

    2018-03-01

    A three-layer-core single-mode large-mode-area fiber is investigated. The three-layer structure in the core, which is composed of a core-index layer, a cladding-index layer, and a depression-index layer, could achieve a large effective area Aeff while maintaining an ultralow bending loss without deteriorating cutoff behaviors. The single-mode large mode area of 100 to 330 μm2 could be achieved in the fiber. The effective area Aeff can be further enlarged by adjusting the layer parameters. Furthermore, the bending property could be improved in this three-layer-core structure. The bending loss could decrease by 2 to 4 orders of magnitude compared with the conventional step-index fiber with the same Aeff. These characteristics of three-layer-core fiber suggest that it can be used in large-mode-area wide-bandwidth high-capacity transmission or high-power optical fiber laser and amplifier in optical communications, which could be used for the basic physical layer structure of big data storage, reading, calculation, and transmission applications.

  17. Multi-dimensional design window search system using neural networks in reactor core design

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki

    2000-02-01

    In the reactor core design, many parametric survey calculations should be carried out to decide an optimal set of basic design parameter values. They consume a large amount of computation time and labor in the conventional way. To support directly design work, we investigate a procedure to search efficiently a design window, which is defined as feasible design parameter ranges satisfying design criteria and requirements, in a multi-dimensional space composed of several basic design parameters. We apply the present method to the neutronics and thermal hydraulics fields and develop the multi-dimensional design window search system using it. The principle of the present method is to construct the multilayer neural network to simulate quickly a response of an analysis code through a training process, and to reduce computation time using the neural network without parametric study using analysis codes. The system works on an engineering workstation (EWS) with efficient man-machine interface for pre- and post-processing. This report describes the principle of the present method, the structure of the system, the guidance of the usages of the system, the guideline for the efficient training of neural networks, the instructions of the input data for analysis calculation and so on. (author)

  18. Design concepts to enhance nuclear power plant protection

    International Nuclear Information System (INIS)

    Ericson, D.M. Jr.; Varnado, G.B.

    1980-01-01

    Using a modern design for a nuclear power plant as a point of departure, this study examines the enhancement of protection which may be achieved by changes to the design. These changes include concepts such as complete physical separation of redundant trains of safety equipment, hardened enclosures for water storage tanks, and hardened shutdown heat removal systems. The degree of enhancement (value) is examined in terms such as the potential reduction in the number of vital areas and the increase in probability of adversary sequence interruption. The impacts considered include constraints imposed upon operations and maintenance personnel and increased capital and operating costs. The study concludes that structural design changes alone do not provide significant increases in protection

  19. Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

    International Nuclear Information System (INIS)

    Papukchiev, Angel; Schaefer, Anselm

    2008-01-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

  20. Status of reactor core design code system in COSINE code package

    International Nuclear Information System (INIS)

    Chen, Y.; Yu, H.; Liu, Z.

    2014-01-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  1. Status of reactor core design code system in COSINE code package

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

    2014-07-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

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

    International Nuclear Information System (INIS)

    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 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 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  4. Science and mathematics teachers’ core teaching conceptions and their implications for engaging in cross-curricular innovations

    Directory of Open Access Journals (Sweden)

    Hanne Møller Andersen

    2010-04-01

    Full Text Available Previous studies have found core teaching conceptions (CTCs to influence teachers’ actions, i.e. how they engage with new teaching practices (e.g. Lotter, Harwood, & Bonner, 2007. This study explores typical CTCs and their subject specific nature in a sample of teachers from physics, biology, and mathematics in Danish upper secondary school. Teachers’ CTCs were investigated through their essay responses to a set of open core questions, administered through a web-platform. Results demonstrate that teachers’ CTCs come in subject specific flavours, encompassing their purpose for teaching the subject, their conceptions of teaching and learning, and their conceptions of interdisciplinary teaching. It is argued that such differences shape teachers’ engagement with new cross-curricular innovations in the Danish context. Assessing and addressing typical and personal CTCs are found to be crucial to a successful implementation of current reform-initiatives, for teacher training, and for self-regulated professional development among teachers.

  5. Current directions in core-shell nanoparticle design

    Science.gov (United States)

    Schärtl, Wolfgang

    2010-06-01

    Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance, affording special types of multi-functionalized and bio-compatible nanoparticles. In this new review article, I try to address the most important developments during the last ten years. This overview is mainly based on frequently cited and more specialized recent review articles from leaders in their respective field. We will consider a variety of nanoscopic core-shell architectures from highly fluorescent nanoparticles (NPs), protected magnetic NPs, multifunctional NPs, thermoresponsive NPs and biocompatible systems to, finally, smart drug-delivery systems.Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance, affording special types of multi-functionalized and bio-compatible nanoparticles. In this new review article, I try to address the most important developments during the last ten years. This overview is mainly based on frequently cited and more specialized recent review articles from leaders in their respective field. We will consider a variety of nanoscopic core-shell architectures from highly fluorescent nanoparticles (NPs), protected magnetic NPs, multifunctional NPs, thermoresponsive NPs and biocompatible systems to, finally, smart drug-delivery systems

  6. Alternative design concept for the second Glass Waste Storage Building

    International Nuclear Information System (INIS)

    Rainisch, R.

    1992-10-01

    This document presents an alternative design concept for storing canisters filled with vitrified waste produced at the Defense Waste Processing Facility (DWPF). The existing Glass Waste Storage Building (GWSB1) has the capacity to store 2,262 canisters and is projected to be completely filled by the year 2000. Current plans for glass waste storage are based on constructing a second Glass Waste Storage Building (GWSB2) once the existing Glass Waste Storage Building (GWSB1) is filled to capacity. The GWSB2 project (Project S-2045) is to provide additional storage capacity for 2,262 canisters. This project was initiated with the issue of a basic data report on March 6, 1989. In response to the basic data report Bechtel National, Inc. (BNI) prepared a draft conceptual design report (CDR) for the GWSB2 project in April 1991. In May 1991 WSRC Systems Engineering issued a revised Functional Design Criteria (FDC), the Rev. I document has not yet been approved by DOE. This document proposes an alternative design for the conceptual design (CDR) completed in April 1991. In June 1992 Project Management Department authorized Systems Engineering to further develop the proposed alternative design. The proposed facility will have a storage capacity for 2,268 canisters and will meet DWPF interim storage requirements for a five-year period. This document contains: a description of the proposed facility; a cost estimate of the proposed design; a cost comparison between the proposed facility and the design outlined in the FDC/CDR; and an overall assessment of the alternative design as compared with the reference FDC/CDR design

  7. Overview of the design of core restraint systems

    International Nuclear Information System (INIS)

    Heinecke, J.

    1984-01-01

    The optimization of the core restraint system is an important condition for the safe and reliable operation of a fast breeder reactor. For KNK II which is under successful operation and SNR 300 all requirements for safety and operation have been met with help of a ring type system. For SNR 2 the decision between the ring type system and the free standing core has to be done in the near future. Within these considerations the advantages of a ring type restraint system of limiting deflections during operation and limiting of possible movements under seismic conditions have to be balanced against the somewhat more complicated structure of the ring type restraint system

  8. Design features affecting dynamic behaviour of fast reactor cores

    International Nuclear Information System (INIS)

    Kayser, G.; Gouriou, A.

    1981-06-01

    The study of dynamic response of an LMFBR to normal and accidental transients needs first of all a simulation code taking into account all the important effects. The DYN-1 code aims at this target. It represents with a sufficiently accurate meshing the core in a 20 geometry for the thermal and reactivity effects, while the kinetics of this core are calculated with a point model. The primary pool, secondary loops, steam generator are also represented, as well as the control and protective systems. We give a short description of this code. Simpler codes are sometimes good enough for parametric studies

  9. CAREM 25: actual status of the core neutronic design. Calculation line

    International Nuclear Information System (INIS)

    Lecot, C.A.

    1990-01-01

    This work follows the one titled 'Criteria for the CAREM 25 reactor core design. Neutronic aspects' presented at this congress, gives in detail the typical values regarding the core defined at this point. Besides, the neutronic calculation line used for the CAREM 25 reactor design is presented. (Author) [es

  10. 75 FR 80571 - Core Principles and Other Requirements for Designated Contract Markets

    Science.gov (United States)

    2010-12-22

    ... Part II Commodity Futures Trading Commission 17 CFR Parts 1, 16, and 38 Core Principles and Other... CFR Parts 1, 16, and 38 RIN 3038-AD09 Core Principles and Other Requirements for Designated Contract... Principles 1. Subpart B--Designation as Contract Market 2. Subpart C--Compliance With Rules i. Proposed Sec...

  11. Evolution of design concepts for remotely maintainable equipment racks

    International Nuclear Information System (INIS)

    Peishel, F.L.; Mouring, R.W.; Schrock, S.L.

    1986-01-01

    Equipment racks have been used to support process equipment in radioactive facilities for many years. Improvements in the design of these racks have evolved relatively slowly primarily as a result of limitations in the capabilities of maintenance equipment; that is, tasks could only be approached from above using bridge cranes with viewing primarily through periscopes. In recent years, however, technological advances have been made by the Consolidated Fuel Reprocessing Program (CFRP) at Oak Ridge National Laboratory (ORNL) in bridge-mounted servomanipulators with onboard auxiliary hoists and television viewing systems. These advances permit full cell coverage by the manipulator arms which, in turn, allow maintenance tasks to be approached horizontally as well as from above. Maintainable equipment items can be stacked vertically on a rack because total overhead access is less important and maintenance tasks that would not have been attempted in the past can now be performed. These advances permit greater flexibility in the design and cell layout of the racks and lead to concepts that could significantly increase the availability of a facility. The evolution of rack design and a description of the alternative concepts based on present maintenance systems capabilities are presented in this paper. 13 refs., 11 figs

  12. Melt cooling by bottom flooding. The COMET core-catcher concept

    International Nuclear Information System (INIS)

    Foit, Jerzy Jan; Alsmeyer, Hans; Tromm, Walter; Buerger, Manfred; Journeau, Christophe

    2009-01-01

    The COMET concept has been developed to cool an ex-vessel corium melt in case of a hypothetical severe accident leading to vessel melt-through. After erosion of a sacrificial concrete layer the melt is passively flooded by bottom injection of coolant water. The open porosities and large surface that are generated during melt solidification form a porous permeable structure that is permanently filled with the evaporating water and thus allows an efficient short-term as well as long-term removal of the decay heat. The advantages of this concept are the fast cool-down and complete solidification of the melt within less than one hour typically. This stops further release of fission products from the corium. A drawback may be the fast release of steam during the quenching process. Several experimental series have been performed by FZK (Germany) to test and optimise the functionality of the different variants of the COMET concept. Thermite generated melts of iron and aluminium oxide were used. The large scale COMET-H test series with sustained inductive heating includes nine experiments performed with an array of water injection channels embedded in a sacrificial concrete layer. Variation of the water inlet pressure and melt height showed that melts up to 50 cm height can be safely cooled with an overpressure of the coolant water of 0.2 bar. The CometPC concept is based on cooling by flooding the melt from the bottom through layers of porous, water filled concrete. The third variant of the COMET design, CometPCA, uses a layer of porous, water filled concrete CometPCA from which flow channels protrude into the layer of sacrificial concrete. This modified concept combines the advantages of the original COMET concept with flow channels and the high resistance of a water-filled porous concrete layer against downward melt attack. Four large scale CometPCA experiments (FZK, Germany) have demonstrated an efficient cooling of melts up to 50 cm height using the recommended water

  13. Thermal hydraulic design of a hydride-fueled inverted PWR core

    International Nuclear Information System (INIS)

    Malen, J.A.; Todreas, N.E.; Hejzlar, P.; Ferroni, P.; Bergles, A.

    2009-01-01

    An inverted PWR core design utilizing U(45%, w/o)ZrH 1.6 fuel (here referred to as U-ZrH 1.6 ) is proposed and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled with U-ZrH 1.6 . The inverted design features circular cooling channels surrounded by prisms of fuel. Hence the relative position of coolant and fuel is inverted with respect to the standard rod bundle design. Inverted core designs with and without twisted tape inserts, used to enhance critical heat flux, were analyzed. It was found that higher power and longer cycle length can be concurrently achieved by the inverted core with twisted tape relative to the optimal standard core, provided that higher core pressure drop can be accommodated. The optimal power of the inverted design with twisted tape is 6869 MW t , which is 135% of the optimally powered standard design (5080 MW t -determined herein). Uncertainties in this design regarding fuel and clad dimensions needed to accommodate mechanical loads and fuel swelling are presented. If mechanical and neutronic feasibility of these designs can be confirmed, these thermal assessments imply significant economic advantages for inverted core designs.

  14. Teaching to the Common Core by Design, Not Accident

    Science.gov (United States)

    Phillips, Vicki; Wong, Carina

    2012-01-01

    The Bill & Melinda Gates Foundation has created tools and supports intended to help teachers adapt to the Common Core State Standards in English language arts and mathematics. The tools seek to find the right balance between encouraging teachers' creativity and giving them enough guidance to ensure quality. They are the product of two years of…

  15. Common Core Standards and their Impact on Standardized Test Design

    NARCIS (Netherlands)

    Polleck, J.N.; Jeffery, J.V.

    2017-01-01

    With adoption of the Common Core (CCSS) in a majority of U.S. states came developmentof new high-stakes exams. Though researchers have investigated CCSS andrelated policies, less attention has been directed toward understanding how standardsare translated into testing. Due to the influence that

  16. Design studies for the Mark-III core of experimental multi-purpose VHTR

    International Nuclear Information System (INIS)

    Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu; Shindo, Ryuiti; Arai, Taketoshi

    1979-08-01

    The Mark-III core in the first conceptual design made in 1975 is a fundamental core for VHTR. Subsequently, further design studies were made fuel loading scheme and control rod withdrawal sequence for the core to increase its safety margin (shutdown margin, etc.) and operational margin (minimum Reynolds number, maximum fuel temperature, etc.). It was shown that the Mark-III should exhibit the performance expected of VHTR, unless changes are made in the preconditions for its nuclear, thermal-hydraulic design. Also, the needs as below were indicated: (1) reasonable core design criteria and guidelines, (2) fuel-loading-scheme requirements in fuel management, fuel misloading and reactor operation, (3) confirmation on precision of the core design method and its further refinement. (author)

  17. Mechanical design

    Science.gov (United States)

    1976-01-01

    Design concepts for a 1000 mw thermal stationary power plant employing the UF6 fueled gas core breeder reactor are examined. Three design combinations-gaseous UF6 core with a solid matrix blanket, gaseous UF6 core with a liquid blanket, and gaseous UF6 core with a circulating blanket were considered. Results show the gaseous UF6 core with a circulating blanket was best suited to the power plant concept.

  18. The EPR-a comprehensive design concept against external events

    International Nuclear Information System (INIS)

    Waas, U.; Stoll, U.

    2005-01-01

    The main objective of design provisions against external hazards is to ensure that the safety functions required to bring the plant to safe shutdown are not inadmissible affected. In the design of the EPR particular attention was paid to the following external hazards: Earthquake, Airplane crash, Explosion pressure wave. The design concept for these events is discussed below. The standard EPR covers a large range of possible site conditions, the design earthquake enveloping safe shutdown earthquakes (SSE) to be expected for potential sites. The Basic Design was developed for the seismic loads given in the European Utility Requirements with a horizontal free field Peak Ground Acceleration (PGA) for all site conditions of 0.25 g. The seismic protection is based on a deterministic design approach, with the intention of ensuring the safety functions in case of SSE. The loads for the design basis airplane crash and - if required - for the design extension airplane crash are defined depending on site specific requirements. For the design basis airplane crash as defined in Finland the safety goals are fulfilled for postulated single failure and preventive maintenance as well as for specific unlikely scenarios with local impacts where one redundant train is assumed to be lost. For the design extension airplane crash no single failure and preventive maintenance are assumed. Reactor building (RB), fuel building (FB), safeguard building (SB) 2 and 3 are protected by design against airplane crash. The common base mat of the RB, FB and SBs ensures global stability. To avoid penetration the wall thickness of the outer building structures of RB, FB and SB2/3 is set at 1.80 m (result of an optimization process). To rule out major induced vibrations due to airplane crash the inner building structures are decoupled from the outer walls. The SB 1 and SB4, the main steam and feedwater valve compartments, the diesel buildings and the service water pump buildings are protected against

  19. Design concepts of definitive disposal for high level radioactive wastes

    International Nuclear Information System (INIS)

    Badillo A, V.E.; Alonso V, G.

    2007-01-01

    It is excessively known the importance about finding a solution for the handling and disposition of radioactive waste of all level. However, the polemic is centered in the administration of high level radioactive waste and the worn out fuel, forgetting that the more important volumes of waste its are generated in the categories of low level wastes or of very low level. Depending on the waste that will be confined and of the costs, several technological modalities of definitive disposition exist, in function of the depth of the confinement. The concept of deep geologic storage, technological option proposed more than 40 years ago, it is a concept of isolation of waste of long half life placed in a deep underground installation dug in geologic formations that are characterized by their high stability and their low flow of underground water. In the last decades, they have registered countless progresses in technical and scientific aspects of the geologic storage, making it a reliable technical solution supported with many years of scientific work carried out by numerous institutions in the entire world. In this work the design concepts that apply some countries for the high level waste disposal that its liberate heat are revised and the different geologic formations that have been considered for the storage of this type of wastes. (Author)

  20. Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

    2015-01-01

    As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

  1. Conceptual design of small-sized HTGR system (3). Core thermal and hydraulic design

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo; Sato, Hiroyuki; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio

    2012-06-01

    The Japan Atomic Energy Agency has started the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the 2030s deployment into developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. As one of the conceptual designs in the first stage, the core thermal and hydraulic design of the power generation and steam supply small-sized HTGR system with a thermal power of 50 MW (HTR50S), which was a reference reactor system positioned as a first commercial or demonstration reactor system, was carried out. HTR50S in the first stage has the same coated particle fuel as HTTR. The purpose of the design is to make sure that the maximum fuel temperature in normal operation doesn't exceed the design target. Following the design, safety analysis assuming a depressurization accident was carried out. The fuel temperature in the normal operation and the fuel and reactor pressure vessel temperatures in the depressurization accident were evaluated. As a result, it was cleared that the thermal integrity of the fuel and the reactor coolant pressure boundary is not damaged. (author)

  2. A solar vehicle based on sustainable design concept

    Energy Technology Data Exchange (ETDEWEB)

    Taha, Z.; Sah, J.M.; Passarella, R.; Ghazilla, R.A.R.; Ahmad, N.; Jen, Y.H.; Khai, T.T.; Kassim, Z.; Hasanuddin, I.; Yunus, M. [Malaya Univ., Kuala Lumpur (Malaysia). Faculty of Engineering, Centre for Product Design and Manufacture

    2009-07-01

    This paper described a newly constructed solar vehicle that was built specifically for the 2009 World Solar Challenge (WSC) using off-the-shelf parts. Researchers at the Centre for Product Design and Manufacture at the University of Malaya designed and built the solar car which uses solar energy to charge its batteries. Although the total investment for this sustainable product concept is small compared to other solar vehicles, the car's performance has met expectations. Most of the electrical and mechanical parts can be recycled and reused after the WSC event. The photovoltaic (PV) and maximum power point trackers (MPPT) can be re-used for home applications. The DC motor and the controller can be attached to a bicycle and the aluminium parts which make-up the main body structure can be recycled. The design will result in nearly zero waste. The study showed that the process of combining mechanical and electrical components is not an easy task, particularly at the design stage because of the specific characteristics and functions of the individual parts. This paper described how readily available, off-the-shelf mechanical and electrical components were integrated for the solar vehicle. The conceptual design and the performance of the prototype were also presented. 11 refs., 5 tabs., 11 figs.

  3. A Dissipative Connector for CLT Buildings: Concept, Design and Testing

    Science.gov (United States)

    Scotta, Roberto; Marchi, Luca; Trutalli, Davide; Pozza, Luca

    2016-01-01

    This paper deals with the conception and characterization of an innovative connection for cross-laminated timber (CLT) panels. The connection is designed to provide an adequate level of dissipative capacity to CLT structures also when realized with large horizontal panels and therefore prone to fragile shear sliding failure. The connector, named X-bracket, has been theorized and designed by means of numerical parametric analyses. Furthermore, its cyclic behavior has been verified with experimental tests and compared to that of traditional connectors. Numerical simulations of cyclic tests of different CLT walls anchored to the foundation with X-brackets were also performed to assess their improved seismic performances. Finally, the analysis of the response of a 6 m × 3 m squat wall demonstrates that the developed connection provides good ductility and dissipation capacities also to shear walls realized with a single CLT panel. PMID:28773265

  4. Implementing Firm Dynamic Capabilities Through the Concept Design Process

    DEFF Research Database (Denmark)

    Nedergaard, Nicky; Jones, Richard

    2011-01-01

    It is well understood that firms operating in highly dynamic and fluid markets need to possess strong dynamic capabilities of sensing (market trajectories), seizing (to capitalise on these trajectories), and transformation (in order to implement sustainable strategies). Less understood is how firms...... actually implement these capabilities. A conceptual model showing how managing concept design processes can help firms systematically develop dynamic capabilities and help bridge the gap between the market-oriented and resource-focused strategic perspectives is presented. By placing this model in a design......-driven innovation perspective three theoretical propositions is derived explicating both the paper’s implementation approach to dynamic capabilities as well as new ways of understanding these capabilities. Concluding remarks are made discussing both the paper’s contribution to the strategic marketing literature...

  5. A Dissipative Connector for CLT Buildings: Concept, Design and Testing.

    Science.gov (United States)

    Scotta, Roberto; Marchi, Luca; Trutalli, Davide; Pozza, Luca

    2016-02-26

    This paper deals with the conception and characterization of an innovative connection for cross-laminated timber (CLT) panels. The connection is designed to provide an adequate level of dissipative capacity to CLT structures also when realized with large horizontal panels and therefore prone to fragile shear sliding failure. The connector, named X-bracket, has been theorized and designed by means of numerical parametric analyses. Furthermore, its cyclic behavior has been verified with experimental tests and compared to that of traditional connectors. Numerical simulations of cyclic tests of different CLT walls anchored to the foundation with X-brackets were also performed to assess their improved seismic performances. Finally, the analysis of the response of a 6 m × 3 m squat wall demonstrates that the developed connection provides good ductility and dissipation capacities also to shear walls realized with a single CLT panel.

  6. Design concept for α-hydrogen-substituted nitroxides.

    Science.gov (United States)

    Amar, Michal; Bar, Sukanta; Iron, Mark A; Toledo, Hila; Tumanskii, Boris; Shimon, Linda J W; Botoshansky, Mark; Fridman, Natalia; Szpilman, Alex M

    2015-02-06

    Stable nitroxides (nitroxyl radicals) have many essential and unique applications in chemistry, biology and medicine. However, the factors influencing their stability are still under investigation, and this hinders the design and development of new nitroxides. Nitroxides with tertiary alkyl groups are generally stable but obviously highly encumbered. In contrast, α-hydrogen-substituted nitroxides are generally inherently unstable and rapidly decompose. Herein, a novel, concept for the design of stable cyclic α-hydrogen nitroxides is described, and a proof-of-concept in the form of the facile synthesis and characterization of two diverse series of stable α-hydrogen nitroxides is presented. The stability of these unique α-hydrogen nitroxides is attributed to a combination of steric and stereoelectronic effects by which disproportionation is kinetically precluded. These stabilizing effects are achieved by the use of a nitroxide co-planar substituent in the γ-position of the backbone of the nitroxide. This premise is supported by a computational study, which provides insight into the disproportionation pathways of α-hydrogen nitroxides.

  7. An integrated software system for core design and safety analyses: Cascade-3D

    International Nuclear Information System (INIS)

    Wan De Velde, A.; Finnemann, H.; Hahn, T.; Merk, S.

    1999-01-01

    The new Siemens program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of the most advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management. (authors)

  8. Design of low-loss and highly birefringent hollow-core photonic crystal fiber

    DEFF Research Database (Denmark)

    Roberts, Peter John; Williams, D.P.; Sabert, H.

    2006-01-01

    A practical hollow-core photonic crystal fiber design suitable for attaining low-loss propagation is analyzed. The geometry involves a number of localized elliptical features positioned on the glass ring that surrounds the air core and separates the core and cladding regions. The size of each...... feature is tuned so that the composite core-surround geometry is antiresonant within the cladding band gap, thus minimizing the guided mode field intensity both within the fiber material and at material / air interfaces. A birefringent design, which involves a 2-fold symmetric arrangement of the features...

  9. The integrated code system CASCADE-3D for advanced core design and safety analysis

    International Nuclear Information System (INIS)

    Neufert, A.; Van de Velde, A.

    1999-01-01

    The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

  10. Design and analysis of EI core structured transverse flux linear reluctance actuator

    OpenAIRE

    FENERCİOĞLU, AHMET; AVŞAR, YUSUF

    2015-01-01

    In this study, an EI core linear actuator is proposed for horizontal movement systems. It is a transverse flux linear switched reluctance motor designed with an EI core structure geometrically. The actuator is configured into three phases and at a 6/4 pole ratio, and it has a stationary active stator along with a sliding passive translator. The stator consists of E cores and the translator consists of I cores. The actuator has a yokeless design because the stator and translator have no back i...

  11. Design of the air-core transformer in spherical tokamak

    International Nuclear Information System (INIS)

    Wang Zhongtian; Jian Guangde; Li Fangzhu; Mao Guoping

    2002-01-01

    An ideal current distribution in the air-core transformer coils is obtained using variation principle. Climbing mountain method is utilized for optimizing the dimension and position of the real coils. Not only can the requirement of minimizing the stray field in the plasma region be guaranteed, but also integer turns for the coil can be realized. The latter brings a significant convenience to engineering

  12. Comparison of design margin for core shroud in between design and construction code and fitness-for-service code

    International Nuclear Information System (INIS)

    Dozaki, Koji

    2007-01-01

    Structural design methods for core shroud of BWR are specified in JSME Design and Construction Code, like ASME Boiler and Pressure Vessel Code Sec. III, as a part of core support structure. Design margins are defined according to combination of the structural design method selected and service limit considered. Basically, those margins in JSME Code were determined after ASME Sec. III. Designers can select so-called twice-slope method for core shroud design among those design methods. On the other hand, flaw evaluation rules have been established for core shroud in JSME Fitness-for-Service Code. Twice-slope method is also adopted for fracture evaluation in that code even when the core shroud contains a flaw. Design margin was determined as structural factors separately from Design and Construction Code. As a natural consequence, there is a difference in those design margins between the two codes. In this paper, it is shown that the design margin in Fitness-for-Service Code is conservative by experimental evidences. Comparison of design margins between the two codes is discussed. (author)

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

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

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

  14. A solar powered wireless computer mouse. Industrial design concepts

    Energy Technology Data Exchange (ETDEWEB)

    Reich, N.H.; Van Sark, W.G.J.H.M.; Alsema, E.A.; Turkenburg, W.C. [Department of Science, Technology and Society, Copernicus Institute, Utrecht University, Heidelberglaan 2, 3584 CS Utrecht (Netherlands); Veefkind, M.; Silvester, S. [Industrial Design Engineering, Delft University of Technology, Landbergstraat 15, 2628 CE Delft (Netherlands)

    2009-02-15

    A solar powered wireless computer mouse (SPM) was chosen to serve as a case study for the evaluation and optimization of industrial design processes of photovoltaic (PV) powered consumer systems. As the design process requires expert knowledge in various technical fields, we assessed and compared the following: appropriate selection of integrated PV type, battery capacity and type, possible electronic circuitries for PV-battery coupling, and material properties concerning mechanical incorporation of PV into the encasing. Besides technical requirements, ergonomic aspects and design aesthetics with respect to good 'sun-harvesting' properties influenced the design process. This is particularly important as simulations show users can positively influence energy balances by 'sun-bathing' the PV mouse. A total of 15 SPM prototypes were manufactured and tested by actual users. Although user satisfaction proved the SPM concept to be feasible, future research still needs to address user acceptance related to product dimensions and user willingness to pro-actively 'sun-bath' PV powered products in greater detail. (author)

  15. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  16. Analysis of a basic core performance for FBR core nuclear design. 3

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    1999-03-01

    The spatial distribution of reaction rates in the ZPPR-13A, having an axially heterogeneous core, has been analyzed. The ZPPR-13A core is treated as a 2-dimensional RZ configuration consisting of a homogeneous core. The analysis is performed by utilizing both probabilistic and deterministic treatments. The probabilistic treatment is performed with the Monte Carlo Code MVP running with continuous energy variable. By comparing the results obtained by both treatments and reviewing the calculation method of effective resonance cross sections, for deterministic treatment, utilized for the reaction rate distributions, it is revealed that the present treatment of effective resonance cross sections is not accurate, since there are observed effects due to dependence on energy group number or energy group width, and on anisotropic scattering. To utilize multi-band method for calculating effective resonance cross sections, widely used by the European researchers, the computer code GROUPIE is installed and the performance of the code is confirmed. Although, in order to improve effective resonance cross sections accuracy, the thermal neutron reactor standard code system SRAC-95 was introduced last year in which the ultra-fine group spectrum calculation module PEACO worked specially under the restriction that number of nuclei having resonance cross section, in any zone, should be less than three, because collision probabilities were obtained by an interpolation method. This year, the module is improved so that these collision probabilities are directly calculated, and by this improvement the highly accurate effective resonance cross sections below the energy of 40.868 keV can be calculated for whole geometrical configurations considered. To extend the application range of the module PEACO, the cross sections of sodium and structure material nuclei are prepared so that they are also represented as ultra-fine group cross sections. By such modifications of cross section library

  17. Mechanical design of core components for a high performance light water reactor with a three pass core

    International Nuclear Information System (INIS)

    Fischer, Kai; Schneider, Tobias; Redon, Thomas; Schulenberg, Thomas; Starflinger, Joerg

    2007-01-01

    Nuclear reactors using supercritical water as coolant can achieve more than 500 deg. C core outlet temperature, if the coolant is heated up in three steps with intermediate mixing to avoid hot streaks. This method reduces the peak cladding temperatures significantly compared with a single heat up. The paper presents an innovative mechanical design which has been developed recently for such a High Performance Light Water Reactor. The core is built with square assemblies of 40 fuel pins each, using wire wraps as grid spacers. Nine of these assemblies are combined to a cluster having a common head piece and a common foot piece. A downward flow of additional moderator water, separated from the coolant, is provided in gaps between the assemblies and in a water box inside each assembly. The cluster head and foot pieces and mixing chambers, which are key components for this design, are explained in detail. (authors)

  18. Design factors affecting dynamic behaviour of fast reactor cores. UK review paper

    Energy Technology Data Exchange (ETDEWEB)

    Brindley, K W [National Nuclear Corporation Ltd., Risley, Warrington (United Kingdom); Perks, M A [United Kingdom Atomic Energy Authority, Risley, Warrington (United Kingdom)

    1982-01-01

    This paper summarises the consideration that has been given in the UK to the following factors that affect the dynamic behaviour of fast reactor cores: fuel design - Pu/u homogeneity, fuel expansion, fuel-clad gaps, uranium fraction. Structural response - CR supports, diagrid, sub-assembly bowing sodium expansion coefficients - low void cores including heterogenous cores. Calculational methods and models are outlined and some experimental results are discussed. (author)

  19. Polymer Design and Processing for Liquid-Core waveguides

    DEFF Research Database (Denmark)

    Sagar, Kaushal Shashikant

    precursor material. Upon attaining thermodynamically stable gyroid phase segregation, nanoporosity is induced by chemically removing PDMS, the so-called sacrificial block. The isotropic nanoporosity in the polymer is utilized in fabricating a novel type of waveguides for opto-fluidic applications, which we...... are spontaneously filled with water by capillary suction, forming the core, while the unmodified hydrophobic regions remain dry, forming the clad. Two types of photo-modification reactions are presented in this thesis: photo-oxidation and thiol-ene photo-clicking. The hydrophilicity is firstly induced by surface...

  20. Addressing pre-service teachers' understandings and difficulties with some core concepts in the special theory of relativity

    International Nuclear Information System (INIS)

    Selcuk, Gamze Sezgin

    2011-01-01

    The aim of this study is to investigate pre-service teachers' understanding of and difficulties with some core concepts in the special theory of relativity. The pre-service teachers (n = 185) from the Departments of Physics Education and Elementary Science Education at Dokuz Eylul University (in Turkey) participated. Both quantitative and qualitative research methods were used in this study. Students' understanding of and difficulties with core elements (time, length, mass and density) were tested using a paper-and-pencil questionnaire (including four questions) and in-depth interviews after the instruction of related modern physics topics. The analyses of the collected data were based on quantitative and qualitative techniques. The results indicate that pre-service teachers at different academic levels have specific and considerable difficulties with proper time, time dilation, proper length, mass and relativistic density concepts. In this paper, the conclusions of the study and implications for physics teaching are discussed.