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.)

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

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)

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)

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

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

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

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)

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)

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)

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)

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.

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

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

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)

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)

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)

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

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)

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)

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)

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

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)

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.

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

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

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)

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)

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…

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

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)

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)

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

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.

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

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.

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.

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)

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

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)

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

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

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)

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

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)

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)

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.

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

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,…

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

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)

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)

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)

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

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)

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

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)

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…

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)

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.

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)

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

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)

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)

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

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

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)

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)

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

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.

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

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

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

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

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

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)

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

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)

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

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)

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

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.

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)

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

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

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)

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

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

Overview of homogeneous versus heterogeneous core configuration trade-off studies

International Nuclear Information System (INIS)

Chang, Y.I.

1982-01-01

The most significant development in core design trend in the U.S. LMFBR program has been the increased attention given to the heterogeneous core design concept. In recent years, numerous core configuration trade-off studies have been carried out to quantify advantages and disadvantages of the heterogeneous concept relative to the homogeneous concept, and a consensus is emerging among the U.S. core designers. It appears that the technical and economic performance differences between homogeneous and heterogeneous core designs are very small; however, the heterogeneous concept provides a definite safety/licensing advantage. he technical and economic performance comparison between homogeneous and heterogeneous core configurations is difficult to quantify. In fact, in most cases, the perceived advantages and/or disadvantages are dictated by the consistency in the comparison (optimized for one concept versus non-optimized for the other, etc.) rather than by any inherent differences. Some of the technical and economic issues relevant to the homogeneous versus heterogeneous comparison are summarized

Overview of homogeneous versus heterogeneous core configuration trade-off studies

Energy Technology Data Exchange (ETDEWEB)

Chang, Y I [Applied Physics Division, Argonne National Laboratory, Argonne, IL (United States)

1982-01-01

The most significant development in core design trend in the U.S. LMFBR program has been the increased attention given to the heterogeneous core design concept. In recent years, numerous core configuration trade-off studies have been carried out to quantify advantages and disadvantages of the heterogeneous concept relative to the homogeneous concept, and a consensus is emerging among the U.S. core designers. It appears that the technical and economic performance differences between homogeneous and heterogeneous core designs are very small; however, the heterogeneous concept provides a definite safety/licensing advantage. he technical and economic performance comparison between homogeneous and heterogeneous core configurations is difficult to quantify. In fact, in most cases, the perceived advantages and/or disadvantages are dictated by the consistency in the comparison (optimized for one concept versus non-optimized for the other, etc.) rather than by any inherent differences. Some of the technical and economic issues relevant to the homogeneous versus heterogeneous comparison are summarized.

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)

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.

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.

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

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)

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

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)

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

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.

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.).

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

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

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

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)

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

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)

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.

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)

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

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

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

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

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.

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)

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

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

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)

Heterogeneous gas core reactor

International Nuclear Information System (INIS)

Han, K.I.

1977-01-01

Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant

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)

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.

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)

Development of inherent core technologies for advanced reactor

International Nuclear Information System (INIS)

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

1999-03-01

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

Development of inherent core technologies for advanced reactor

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

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

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.

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…

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).

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

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...

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.

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

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

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)

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)

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

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.

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)

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

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

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)

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.

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

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)

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

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.

The EFR project: core and fuel

International Nuclear Information System (INIS)

Francillon, E.; Barnes, D.W.; Pay, A.; Wehmann, U.

1991-01-01

The draft studies on EFR has beginning, in March 1988. The first part of the summary draft has consisted in the core and fuel domains to harmonize the different approaches used in national projects (SPX2-SNR2-CDFR). Rapidly, the core First Consistent Design has been defined with references to the anterior conceptions. Since this definition, studies have been engaged on the management (mean burnup amelioration) and on the conception (breeding gain, sodium void coefficient reduction). After a presentation of the basis options and on the general conception of the fuel assemblies we make a point on the core and the interfaces with the fuel cycle [fr

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.

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...

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

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

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)

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.

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

Wire core reactor for NTP

International Nuclear Information System (INIS)

Harty, R.B.

1991-01-01

The development of the wire core system for Nuclear Thermal Propulsion (NTP) that took place from 1963 to 1965 is discussed. A wire core consists of a fuel wire with spacer wires. It's an annular flow core having a central control rod. There are actually four of these, with beryllium solid reflectors on both ends and all the way around. Much of the information on the concept is given in viewgraph form. Viewgraphs are presented on design details of the wire core, the engine design, engine weight vs. thrust, a technique used to fabricate the wire fuel element, and axial temperature distribution

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

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…

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)

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

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

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.

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

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.

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

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)

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.

Design and evaluation of low-cost laminated wood composite blades for intermediate size wind turbines: Blade design, fabrication concept, and cost analysis

Science.gov (United States)

Lieblein, S.; Gaugeon, M.; Thomas, G.; Zueck, M.

1982-01-01

As part of a program to reduce wind turbine costs, an evaluation was conducted of a laminated wood composite blade for the Mod-OA 200 kW wind turbine. The effort included the design and fabrication concept for the blade, together with cost and load analyses. The blade structure is composed of laminated Douglas fir veneers for the primary spar and nose sections, and honeycomb cored plywood panels for the trailing edges sections. The attachment of the wood blade to the rotor hub was through load takeoff studs bonded into the blade root. Tests were conducted on specimens of the key structural components to verify the feasibility of the concept. It is concluded that the proposed wood composite blade design and fabrication concept is suitable for Mod-OA size turbines (125-ft diameter rotor) at a cost that is very competitive with other methods of manufacture.

Gas core nuclear thermal rocket engine research and development in the former USSR

International Nuclear Information System (INIS)

Koehlinger, M.W.; Bennett, R.G.; Motloch, C.G.; Gurfink, M.M.

1992-09-01

Beginning in 1957 and continuing into the mid 1970s, the USSR conducted an extensive investigation into the use of both solid and gas core nuclear thermal rocket engines for space missions. During this time the scientific and engineering. problems associated with the development of a solid core engine were resolved. At the same time research was undertaken on a gas core engine, and some of the basic engineering problems associated with the concept were investigated. At the conclusion of the program, the basic principles of the solid core concept were established. However, a prototype solid core engine was not built because no established mission required such an engine. For the gas core concept, some of the basic physical processes involved were studied both theoretically and experimentally. However, no simple method of conducting proof-of-principle tests in a neutron flux was devised. This report focuses primarily on the development of the. gas core concept in the former USSR. A variety of gas core engine system parameters and designs are presented, along with a summary discussion of the basic physical principles and limitations involved in their design. The parallel development of the solid core concept is briefly described to provide an overall perspective of the magnitude of the nuclear thermal propulsion program and a technical comparison with the gas core concept

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

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.

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

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

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.

Introduction to Open Core Protocol Fastpath to System-on-Chip Design

CERN Document Server

Schwaderer, W David

2012-01-01

This book introduces Open Core Protocol (OCP), not as a conventional hardware communications protocol but as a meta-protocol: a means for describing and capturing the communications requirements of an IP core, and mapping them to a specific set of signals with known semantics.  Readers will learn the capabilities of OCP as a semiconductor hardware interface specification that allows different System-On-Chip (SoC) cores to communicate.  The OCP methodology presented enables intellectual property designers to design core interfaces in standard ways. This facilitates reusing OCP-compliant cores across multiple SoC designs which, in turn, drastically reduces design times, support costs, and overall cost for electronics/SoCs. Provides a comprehensive introduction to Open Core Protocol, which is more accessible than the full specification; Designed as a hands-on, how-to guide to semiconductor design; Includes numerous, real “usage examples” which are not available in the full specification; Integrates coverag...

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.)

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)

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

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.

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)

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...

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.

Design of a New Research Reactor: Preliminary Conceptual Design (3rd Year)

International Nuclear Information System (INIS)

Park, Cheol; Lee, B. C.; Chae, H. T. and others

2006-01-01

A research reactor design is a kind of integral engineering project and a process to obtain a concrete shape through several years of concept development, conceptual design, basic design and detail design. So it requires close cooperation in various areas as well as lots of manpower and cost. The overall process at each stage may be said to be similar except for some stage-specific works. In 2005 as last year of a concept development stage, investigations on the various concepts of the fuel, reactor structure and systems which can meet the requirements established. The requirements for the process systems and I and C systems have also been embodied. The major tasks planned at the early of 2005 have been performed for each area of reactor design as follows: Establishment of the fuel and reactor core concept, and the core analysis, Preliminary thermal-hydraulic and safety analyses for the conceptual cores, Establishment and improvement of analysis system, Concept developments of the reactor structures and major systems, Test and test plan to verify the developed concepts, International cooperation to establish the foundations for exporting a research reactor

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...

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)

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

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…

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

Study on plant concept for gas cooled fast reactor

International Nuclear Information System (INIS)

Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

2003-05-01

In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

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

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…

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

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)

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.

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)

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

Increasing the neutron flux study for the TRR-II core design

International Nuclear Information System (INIS)

Chen, C.-H.; Yang, J.-T.; Chou, Y.-C.

1999-01-01

The maximum unperturbed thermal flux of the originally proposed core design, which is a 6x6 square arrangement with power level of 20 MW and has been presented at the 6th Meeting of IGORR, for the TRR-II reactor is about 2.0x10 14 n/cm 2 -sec. However, it is no longer satisfied the user's requirement, that is, it must reach at least 2.5x10 14 n/cm 2 -sec. In order to enhance the thermal neutron flux, one of the most effective ways is to increase the average power density. Therefore, two new designs with more compact cores are then proposed and studied. One is 5x6 rectangular arrangement with power of 20 MW; the other one is 5x5 square arrangement with power of 16 MW. It is for sure that both core designs can satisfy thermal hydraulic safety limits. The designed parameters related to neutronics are listed and compared fundamentally. According to our calculation, although both cores have similar average power density, the results show that the 5x6/20 MW design has the maximum unperturbed thermal flux in the D 2 O region about 2.7x10 14 n/cm 2 -sec, and the 5x5/16 MW design has 2.5x10 14 n/cm 2 -sec. The maximum thermal flux in the neighborhood of the longer side of the 5x6 core is about 7% higher than the one in the neighborhood of any side of the 5x5 core. This 'long-side effect' gives the 5x6/20 MW core design an advantage of the utilization of the thermal neutron flux in the D 2 O region. In addition, the 5x5 core is also more sensitive to the reactivity change on account of in-core irradiation test facilities. Therefore, under overall considerations the 5x6/20 MW core design is chosen for further detailed design. (author)

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.

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)

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

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....

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

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...

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)

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)

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.

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

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

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

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

Design study of nuclear power systems for deep space explorers. (2) Electricity supply capabilities of solid cores

International Nuclear Information System (INIS)

Yamaji, Akifumi; Takizuka, Takakazu; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime

2009-01-01

This study has been carried out in series with the other study, 'Criticality of Low Enriched Uranium Fueled Core' to explore the possibilities of a solid reactor electricity generation system for supplying propulsion power of a deep space explorer. The design ranges of two different systems are determined with respect to the electric power, the radiator mass, and the operating temperatures of the heat-pipes and thermoelectric converters. The two systems are the core surface cooling with heat-pipe system (CSHP), and the core direct cooling with heat-pipe system (CDHP). The evaluated electric powers widely cover the 1 to 100 kW range, which had long been claimed to be the range that lacked the power sources in space. Therefore, the concepts shown by this study may lead to a breakthrough of the human activities in space. The working temperature ranges of the main components, namely the heat-pipes and thermoelectric converters, are wide and covers down to relatively low temperatures. This is desirable from the viewpoints of broadening the choices, reducing the development needs, and improving the reliabilities of the devices. Hence, it is advantageous for an early establishment of the concept. (author)

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.

Training reactor deployment. Advanced experimental course on designing new reactor cores

International Nuclear Information System (INIS)

Skoda, Radek

2009-01-01

Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

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

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…

Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

International Nuclear Information System (INIS)

Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J.; Flanery, B.K.; Shinko, M.S.; Giffen, R.H.; Rampolla, D.S.

1979-12-01

Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used

NASA's Gravitational - Wave Mission Concept Study

Science.gov (United States)

Stebbins, Robin; Jennrich, Oliver; McNamara, Paul

2012-01-01

With the conclusion of the NASA/ESA partnership on the Laser Interferometer Space Antenna (LISA) Project, NASA initiated a study to explore mission concepts that will accomplish some or all of the LISA science objectives at lower cost. The Gravitational-Wave Mission Concept Study consisted of a public Request for Information (RFI), a Core Team of NASA engineers and scientists, a Community Science Team, a Science Task Force, and an open workshop. The RFI yielded were 12 mission concepts, 3 instrument concepts and 2 technologies. The responses ranged from concepts that eliminated the drag-free test mass of LISA to concepts that replace the test mass with an atom interferometer. The Core Team reviewed the noise budgets and sensitivity curves, the payload and spacecraft designs and requirements, orbits and trajectories and technical readiness and risk. The Science Task Force assessed the science performance by calculating the horizons. the detection rates and the accuracy of astrophysical parameter estimation for massive black hole mergers, stellar-mass compact objects inspiraling into central engines. and close compact binary systems. Three mission concepts have been studied by Team-X, JPL's concurrent design facility. to define a conceptual design evaluate kt,y performance parameters. assess risk and estimate cost and schedule. The Study results are summarized.

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)

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

Integral fast reactor concept

International Nuclear Information System (INIS)

Chang, Y.I.; Marchaterre, J.F.; Sevy, R.H.

1984-01-01

Key features of the IFR consist of a pool-type plant arrangement, a metal fuel-based core design, and an integral fuel cycle with colocated fuel cycle facility. Both the basic concept and the technology base have been demonstrated through actual integral cycle operation in EBR-II. This paper discusses the inherent safety characteristics of the IFR concept

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.

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

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)

Advantages of iron core in a tokamak

International Nuclear Information System (INIS)

Bettis, E.S.; Ballou, J.K.; Becraft, W.R.; Peng, Y.K.M.; Watts, H.L.

1977-01-01

A quantitative comparison of the iron core vs air core concepts was carried out on a preliminary basis by using a representative tokamak reactor design with the following self-consistent reference parameters. In the area of plasma engineering, poloidal field and MHD equilibrium considerations with an unsaturated iron core is discussed. The question of proper poloidal field coils to maintain D-shaped plasmas of relatively high anti β (7%) with a saturated iron core is also discussed. Estimates of the required iron core size, volt seconds, magnetic flux and its influence on force loading on the superconducting toroidal field coils are shown. Conceptual designs of the mechanical structure of an iron core device are presented. Favorable impacts on the OH power supply cost and complexity are indicated

Status review of large fast reactor core designs and their dynamics related features

International Nuclear Information System (INIS)

Spenke, H.; Kiefhaber, E.

1982-01-01

Since several years conventional and unconventional concepts of large fast reactor cores have been investigated in the Federal Republic of Germany at INTERATOM and Kernforschungszentrum Karlsruhe. The work was performed jointly with Belgonucleaire (Belgium). Basically, the studies were aimed at the determination of the performance potential of different core concepts for large fast reactors. Thus the following points were considered: power distribution, neutron fluence and residence time, doubling time, uranium ore consumption, dynamics and safety related features, economics, cooling strategy, core element bowing behaviour. In this paper, the state of the analysis will be presented with emphasis on those points relevant for this meeting. However, we have to make clear, that dynamic and accident studies are still under way and that we are not yet able to cover these aspects in a quantitative manner. This is due to the fact, that the efforts in the DeBeNe-countries have been concentrated on the work necessary for being granted the different licenses for SNR 300, fast breeder prototype reactor near Kalkar. As we expect to obtain these important licenses at the beginning of 1982, an increased man power can be devoted to studies of dynamic and safety problems of large fast cores from that time on. These studies have to fit into the planning recently announced by the utility ESK who will be ordering SNR 2, the first demonstration breeder reactor of Germany, Belgium, Netherlands and France. The planning calls for concept decisions in 1983, leading to an engineering contract for SNR 2 in 1983/1984. Accordingly we shall have to complete and evaluate the ongoing core concept Investigations till 1983 resulting in a subsequent final choice

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)

Optimization method development of the core characteristics of a fast reactor in order to explore possible high performance solutions (a solution being a consistent set of fuel, core, system and safety)

International Nuclear Information System (INIS)

Ingremeau, J.-J.X.

2011-01-01

In the study of any new nuclear reactor, the design of the core is an important step. However designing and optimising a reactor core is quite complex as it involves neutronics, thermal-hydraulics and fuel thermomechanics and usually design of such a system is achieved through an iterative process, involving several different disciplines. In order to solve quickly such a multi-disciplinary system, while observing the appropriate constraints, a new approach has been developed to optimise both the core performance (in-cycle Pu inventory, fuel burn-up, etc...) and the core safety characteristics (safety estimators) of a Fast Neutron Reactor. This new approach, called FARM (Fast Reactor Methodology) uses analytical models and interpolations (Meta-models) from CEA reference codes for neutronics, thermal-hydraulics and fuel behaviour, which are coupled to automatically design a core based on several optimization variables. This global core model is then linked to a genetic algorithm and used to explore and optimise new core designs with improved performance. Consideration has also been given to which parameters can be best used to define the core performance and how safety can be taken into account.This new approach has been used to optimize the design of three concepts of Gas cooled Fast Reactor (GFR). For the first one, using a SiC/SiCf-cladded carbide-fuelled helium-bonded pin, the results demonstrate that the CEA reference core obtained with the traditional iterative method was an optimal core, but among many other possibilities (that is to say on the Pareto front). The optimization also found several other cores which exhibit some improved features at the expense of other safety or performance estimators. An evolution of this concept using a 'buffer', a new technology being developed at CEA, has hence been introduced in FARM. The FARM optimisation produced several core designs using this technology, and estimated their performance. The results obtained show that

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

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

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.

[Three-dimensional computer aided design for individualized post-and-core restoration].

Science.gov (United States)

Gu, Xiao-yu; Wang, Ya-ping; Wang, Yong; Lü, Pei-jun

2009-10-01

To develop a method of three-dimensional computer aided design (CAD) of post-and-core restoration. Two plaster casts with extracted natural teeth were used in this study. The extracted teeth were prepared and scanned using tomography method to obtain three-dimensional digitalized models. According to the basic rules of post-and-core design, posts, cores and cavity surfaces of the teeth were designed using the tools for processing point clouds, curves and surfaces on the forward engineering software of Tanglong prosthodontic system. Then three-dimensional figures of the final restorations were corrected according to the configurations of anterior teeth, premolars and molars respectively. Computer aided design of 14 post-and-core restorations were finished, and good fitness between the restoration and the three-dimensional digital models were obtained. Appropriate retention forms and enough spaces for the full crown restorations can be obtained through this method. The CAD of three-dimensional figures of the post-and-core restorations can fulfill clinical requirements. Therefore they can be used in computer-aided manufacture (CAM) of post-and-core restorations.

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.

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.

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

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.

CONSUL code package application for LMFR core calculations

Energy Technology Data Exchange (ETDEWEB)

Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)

2008-07-01

CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)

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.

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.)

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

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

Plant Control Concept for the Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Kim, Eui Kwang; Kim, S. O.

2010-12-01

A power plant is designed for incorporation into a utility's grid system and follows the load demand through the steam generator, intermediate heat exchanger(IHX), from the nuclear core. During the load-following transients, various plant parameters must be controlled to protect the reactor core and other components in the plant. The purpose of this report is to review design considerations to establish SFR plant control and to design plant control concepts. The governing equations and solution procedure of the computer code to calculate plant temperature conditions during the part-load operation was reviewed and 4 types of plant operation concepts were designed, and the results of the calculations were compared

A reverse depletion method for pressurized water reactor core reload design

International Nuclear Information System (INIS)

Downar, T.J.; Kin, Y.J.

1986-01-01

Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. The large numbers of burnable poison pins used to control the power peaking at the in-board fresh fuel positions have introduced 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 burnable poison pins in the fresh fuel. A new method for performing core design more suitable for low-leakage fuel management is reported. A procedure was developed that uses the wellknown ''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 burnable poison 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 provided for the maximum utility to PWR core reload design

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.

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...

Simulation of the Long period Core Design for WH type of KHNP

Energy Technology Data Exchange (ETDEWEB)

Jung, Ji-Eun; Moon, Sang-Rae [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

2016-10-15

The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve.

Simulation of the Long period Core Design for WH type of KHNP

International Nuclear Information System (INIS)

Jung, Ji-Eun; Moon, Sang-Rae

2016-01-01

The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve

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

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)

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

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

The whiteStar development project: Westinghouse's next generation core design simulator and core monitoring software to power the nuclear renaissance

International Nuclear Information System (INIS)

Boyd, W. A.; Mayhue, L. T.; Penkrot, V. S.; Zhang, B.

2009-01-01

The WhiteStar project has undertaken the development of the next generation core analysis and monitoring system for Westinghouse Electric Company. This on-going project focuses on the development of the ANC core simulator, BEACON core monitoring system and NEXUS nuclear data generation system. This system contains many functional upgrades to the ANC core simulator and BEACON core monitoring products as well as the release of the NEXUS family of codes. The NEXUS family of codes is an automated once-through cross section generation system designed for use in both PWR and BWR applications. ANC is a multi-dimensional nodal code for all nuclear core design calculations at a given condition. ANC predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. BEACON is an advanced core monitoring and support system which uses existing instrumentation data in conjunction with an analytical methodology for on-line generation and evaluation of 3D core power distributions. This new system is needed to design and monitor the Westinghouse AP1000 PWR. This paper describes provides an overview of the software system, software development methodologies used as well some initial results. (authors)

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.

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.)

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

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.

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.

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

Baseline Design Compliance Matrix for the Rotary Mode Core Sampling System

International Nuclear Information System (INIS)

LECHELT, J.A.

2000-01-01

The purpose of the design compliance matrix (DCM) is to provide a single-source document of all design requirements associated with the fifteen subsystems that make up the rotary mode core sampling (RMCS) system. It is intended to be the baseline requirement document for the RMCS system and to be used in governing all future design and design verification activities associated with it. This document is the DCM for the RMCS system used on Hanford single-shell radioactive waste storage tanks. This includes the Exhauster System, Rotary Mode Core Sample Trucks, Universal Sampling System, Diesel Generator System, Distribution Trailer, X-Ray Cart System, Breathing Air Compressor, Nitrogen Supply Trailer, Casks and Cask Truck, Service Trailer, Core Sampling Riser Equipment, Core Sampling Support Trucks, Foot Clamp, Ramps and Platforms and Purged Camera System. Excluded items are tools such as light plants and light stands. Other items such as the breather inlet filter are covered by a different design baseline. In this case, the inlet breather filter is covered by the Tank Farms Design Compliance Matrix

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

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.)

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...

Development of high performance core for large fast breeder reactors

International Nuclear Information System (INIS)

Inoue, Kotaro; Kawashima, Katsuyuki; Watari, Yoshio.

1982-01-01

Subsequently to the fast breeder prototype reactor ''Monju'', the construction of a demonstration reactor with 1000 MWe output is planned. This research aims at the establishment of the concept of a large core with excellent fuel breeding property and safety for a demonstration and commercial reactors. For the purpose, the optimum specification of fuel design as a large core was clarified, and the new construction of a core was examined, in which a disk-shaped blanket with thin peripheral edge is introduced at the center of a core. As the result, such prospect was obtained that the time for fuel doubling would be 1/2, and the energy generated in a core collapse accident would be about 1/5 as compared with a large core using the same fuel as ''Monju''. Generally, as a core is enlarged, the rate of breeding lowers. If a worst core collapse accident occurs, the scale of accident will be very large in the case of a ''Monju'' type large core. In an unhomogeneous core, an internal blanket is provided in the core for the purpose of improving the breeding property and safety. Hitachi Ltd. developed the concept of a large core unhomogeneous in axial direction and proposed it. The research on the fuel design for a large core, an unhomogeneous core and its core collapse accident is reported. (Kako, I.)

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

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.

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

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

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

Tuned Chamber Core Panel Acoustic Test Results

Science.gov (United States)

Schiller, Noah H.; Allen, Albert R.

2016-01-01

This report documents acoustic testing of tuned chamber core panels, which can be used to supplement the low-frequency performance of conventional acoustic treatment. The tuned chamber core concept incorporates low-frequency noise control directly within the primary structure and is applicable to sandwich constructions with a directional core, including corrugated-, truss-, and fluted-core designs. These types of sandwich structures have long, hollow channels (or chambers) in the core. By adding small holes through one of the facesheets, the hollow chambers can be utilized as an array of low-frequency acoustic resonators. These resonators can then be used to attenuate low-frequency noise (below 400 Hz) inside a vehicle compartment without increasing the weight or size of the structure. The results of this test program demonstrate that the tuned chamber core concept is effective when used in isolation or combined with acoustic foam treatments. Specifically, an array of acoustic resonators integrated within the core of the panels was shown to improve both the low-frequency absorption and transmission loss of the structure in targeted one-third octave bands.

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

New Sodium Cooled Long-Life Cores with Axially Multi-Driver Regions

International Nuclear Information System (INIS)

Hyun, Hae Ri; Hong, Ser Gi

2014-01-01

In this concept of long-life core (they are sometimes called B-B (Breed and Burn)), tall blanket is placed above the relatively short driver fuel. In the initial stage of burning, the power by fission is mostly generated in the driver region and it moves into the blanket region. The power and flux distributions that are highly peaked in the axial direction propagates slowly from the driver into the blanket region. This concept of long-life core fully utilizes the breeding of blanket in the fast spectra and it can achieve very high burnup of fuel. In this work, we introduce new sodium cooled longlife cores rating 600MWe (1800MWt). In these cores, the driver regions are heterogeneously placed into blanket region so as to achieve stabilized and less peaked axial power distribution as depletion proceeds. At present, our study is focused on only two axial driver regions but this concept can be easily extended onto the multi-driver region concept. The cores designed in this paper have two axial driver regions so as to have stabilized and less peaked axial power distributions as depletion proceeds. The results of the core design and analyses show that the cores have very long-lives longer than -49EFPYs and high discharge burnup higher than 200GWD/kg. Additionally, we considered a long-life core having no blanket. As expected, it was shown that these cores have stabilized and less peaked axial power distribution as the fuel depletes. However, the study shows that the cores having two driver regions still show high initial peaking of the axial power distributions and the core can be optimized by changing the driver fuel height

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

Application of the lines of protection concept to the HTGR-SC/C

International Nuclear Information System (INIS)

1981-09-01

This study of the application of the line of protection (LOP) concept to high temperature gas-cooled reactors (HTGRs) was motivated by a desire to develop a simple and straightforward HTGR safety concept that embodies many of the more complicated and seemingly conflicting concepts facing nuclear industry safety today. These concepts include: (1) defense in depth; (2) design basis events; (3) core damage events (degraded cores); (4) probabilistic analysis and risk assessment; (5) numerical safety goals; and (6) plant investment protection. The LOP concept described herein attempts to incorporate many of the important principles of each into a cohesive framework which provides an overall logic, meaning, and direction for conducting HTGR design and research activities

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.

Preliminary Study for Conceptual Design of Advanced Long Life Small Modular Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, T. K. [Argonne National Laboratory, Argonne (United States)

2015-05-15

As one of the non-water coolant Small-Modular Reactor (SMR) core concepts for use in the mid- to long-term, ANL has proposed a 100 MWe Advanced sodium-cooled Fast Reactor core concept (AFR-100) targeting a small grid, transportable from pre-licensed factories to the remote plant site for affordable supply. Various breed-and-burn core concepts have been proposed to extend the reactor cycle length, which includes CANDLE with a cigar-type depletion strategy, TerraPower reactors with fuel shuffling for effective breeding, et al. UNIST has also proposed an ultra-long cycle fast reactor (UCFR) core concept having the power rating of 1000 MWe. By adopting the breed-and-burn strategies, the UCFR core can maintain criticality for a targeting reactor lifetime of 60 years without refueling. The objective of this project is to develop an advanced long-life SMR core concept by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. A conceptual design of long life small modular fast reactor is under development by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. The feasibility of the long-life fast reactor concepts was reviewed to obtain the core design guidelines and the reactor design requirements of long life small modular fast reactor were proposed in this study.

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.

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

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

Neutronic feasibility of an LMFBR super long-life core (SLLC)

International Nuclear Information System (INIS)

Kawashima, Masatoshi; Aoki, Katsutada; Arie, Kazuo; Tsuboi, Yasushi

1988-01-01

The LMFBR Super Long-Life Core (SLLC) concept has evolved over the last few years as one of the targets of innovative approaches for future FBR cost reduction. An idea for SLLC has been developed wherein the core lifetime is extended up to the plant life of about 30 years by applying the radially and axially multi-zoned core concept (the improved homogeneous core concept). The main purpose of the present study is placed on the evaluation of neutronic feasibility of the 1000 MWe class SLLC concept. The core size of the present SLLC, which is approximately 3 to 4 times as large as those of the current 1000 MWe core design, was determined by the limit of the maximum fast neutron fluence level, which was tentatively assumed to be 5-6x10 23 nvt as the target of the future development of advanced cladding materials. Emphasis is placed on the discussion of neutronic performances of cores with oxide fuels rather than metal or carbide fuels. The present study has shown that proper zoning of the different plutonium enrichment fuels at the initial core makes it possible to achieve small enough reactivity loss during 30-year burnup while satisfying mild variation of the subassembly power distributions using a higher fuel volume fraction of about 50%. Effects of important neutronic parameters on the core performances are also discussed. (orig.)

Design of a reactor core in the Oma Full MOX-ABWR

International Nuclear Information System (INIS)

Hama, Teruo

1999-01-01

The Electric Power Development Co., Ltd. has progressed a construction plan on an improved boiling-water reactor aiming at loading of MOX fuel in all reactor cores (full MOX-ABWR) at Oma-cho, Aomori prefecture, which is a last stage on application of approval on establishment at present. Here were described on outlines of reactor core in the full MOX-ABWR and its safety evaluation. For the full MOX-ABWR loading MOX fuel assembly into all reactor core, thermal and mechanical design analysis of fuel bars and core design analysis were conducted. As a result, it was confirmed that judgement standards in mixed core of MOX fuel and uranium fuel were also applicable as well as that in uranium fuel. (G.K.)

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

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

Accident tolerant high-pressure helium injection system concept for light water reactors

International Nuclear Information System (INIS)

Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

2016-01-01

Highlights: • Potential helium injection strategy is proposed for LWR accident scenarios. • Multiple injection sites are proposed for current LWR designs. • Proof-of-concept experimentation illustrates potential helium injection benefits. • Computational studies show an increase in pressure vessel blowdown time. • Current LOCA codes have the capability to include helium for feasibility calculations. - Abstract: While the design of advanced accident-tolerant fuels and structural materials continues to remain the primary focus of much research and development pertaining to the integrity of nuclear systems, there is a need for a more immediate, simple, and practical improvement in the severe accident response of current emergency core cooling systems. Current blowdown and reflood methodologies under accident conditions still allow peak cladding temperatures to approach design limits and detrimentally affect the integrity of core components. A high-pressure helium injection concept is presented to enhance accident tolerance by increasing operator response time while maintaining lower peak cladding temperatures under design basis and beyond design basis scenarios. Multiple injection sites are proposed that can be adapted to current light water reactor designs to minimize the need for new infrastructure, and concept feasibility has been investigated through a combination of proof-of-concept experimentation and computational modeling. Proof-of-concept experiments show promising cooling potential using a high-pressure helium injection concept, while the developed choked-flow model shows core depressurization changes with added helium injection. Though the high-pressure helium injection concept shows promise, future research into the evaluation of system feasibility and economics are needed.Classification: L. Safety and risk analysis

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

Integrated CFD investigation of heat transfer enhancement using multi-tray core catcher in SFR

International Nuclear Information System (INIS)

Rakhi; Sharma, Anil Kumar; Velusamy, K.

2017-01-01

Highlights: • Heat transfer enhancement using multi-tray core catcher for SFR is investigated. • The capability of a single core collector tray is estimated. • Double and triple collector trays with innovative designs is discussed. • Provision of openings in the trays contributed to enhanced natural circulation. - Abstract: To render future SFR more robust and safe, certain BDBE have been considered in the recent years. A Core Disruptive Accident leading to a whole core meltdown scenario has gained the interest of researchers. Various design concepts and safety measures have been suggested and incorporated in design to address such a low probability scenario. A core catcher concept, in particular, has proved to be inevitable as an in-vessel core retention device in SFR for safe retention of core debris arising out after the severe accident. This study aims to analyse the cooling capability of the innovative design concept of core catcher to remove decay heat of degraded core after the accident. First, the capability of single collection tray is established and then the study is extended to two and three collection trays with different design concepts. Transient forms of governing equations of mass, momentum and energy conservations along with k-ε turbulence model are solved by finite volume based CFD solver. Boussinesq approximation is invoked to model buoyancy in sodium. The study shows that a single collection tray is capable of removing up to 20 MW decay heat load in a typical 500 MWe pool type SFR. Further, studies are carried out to improve the natural circulation of sodium around the source, in the lower plenum and to distribute core debris of the whole core to multiple collection trays. It is found that the double and triple collection trays can accommodate decay loads up to 29 MW. Provision of openings in the collection trays has proved to be effective in improving the heat transfer and sodium flow as well as in distributing the core debris to the

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.

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...

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)

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)

Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

Energy Technology Data Exchange (ETDEWEB)

Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

1985-07-01

This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

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

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.

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.

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)

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...

Advanced PWR Core Design with Siemens High-Plutonium-Content MOX Fuel Assemblies

International Nuclear Information System (INIS)

Dieter Porsch; Gerhard Schlosser; Hans-Dieter Berger

2000-01-01

The Siemens experience with plutonium recycling dates back to the late 1960s. Over the years, extensive research and development programs were performed for the qualification of mixed-oxide (MOX) technology and design methods. Today's typical reload enrichments for uranium and MOX fuel assemblies and modern core designs have become more demanding with respect to accuracy and reliability of design codes. This paper presents the status of plutonium recycling in operating high-burnup pressurized water reactor (PWR) cores. Based on actual examples, it describes the validation status of the design methods and stresses current and future needs for fuel assembly and core design including those related to the disposition of weapons-grade plutonium

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...

Fast Flux Test Facility core system

International Nuclear Information System (INIS)

Ethridge, J.L.; Baker, R.B.; Leggett, R.D.; Pitner, A.L.; Waltar, A.E.

1990-11-01

A review of Liquid Metal Reactor (LMR) core system accomplishments provides an excellent road map through the maze of issues that faced reactor designers 10 years ago. At that time relatively large uncertainties were associated with fuel pin and fuel assembly performance, irradiation of structural materials, and performance of absorber assemblies. The extensive core systems irradiation program at the US Department of Energy's Fast Flux Test Facility (FFTF) has addressed each of these principal issues. As a result of the progress made, the attention of long-range LMR planners and designers can shift away from improving core systems and focus on reducing capital costs to ensure the LMR can compete economically in the 21st century with other nuclear reactor concepts. 3 refs., 6 figs., 1 tab

Computer Science Concept Inventories: Past and Future

Science.gov (United States)

Taylor, C.; Zingaro, D.; Porter, L.; Webb, K. C.; Lee, C. B.; Clancy, M.

2014-01-01

Concept Inventories (CIs) are assessments designed to measure student learning of core concepts. CIs have become well known for their major impact on pedagogical techniques in other sciences, especially physics. Presently, there are no widely used, validated CIs for computer science. However, considerable groundwork has been performed in the form…

Design and safety studies on an EFIT core with CERMET fuel

International Nuclear Information System (INIS)

Chen, Xue-Nong; Rineiski, Andrei; Liu, Ping; Maschek, Werner; Matzerath Boccaccini, Claudia; Gabrielli, Fabrizio; Sobolev, Vitaly

2008-01-01

Within the EUROTRANS Programme a European Facility for Industrial Transmutation (EFIT) is under development. This paper deals with the design and safety analyses of an EFIT core with Mo-matrix based CERMET fuel. A three zone core design was developed, which satisfies the EFIT general and specific requirements. The fuel/matrix ratio in each zone is determined for a suitable subcritical level at a k eff of about 0.97 and a total form factor around 1.5. The Pu/MA ratio also determines the transmutation rate and the burn-up characteristics, ranging between 46/54 at% to 40/60 at% for optimizing the reactivity swing and the MA transmutation efficiency. Based on the preliminary core design, safety calculations are performed with SIMMER-III for three types of transient: the unprotected loss of flow (ULOF), the unprotected transient of over power (UTOP) and the unprotected blockage accident (UBA). It can be shown that in the CERMET core the fuel and clad design limits are not violated under the conditions of ULOF and UTOP. In the UBA case, pin failures will happen and lead to a local voiding and reactivity insertion, but a fuel sweep-out process leads to a power reduction and restricts the core degradation. (authors)

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)

RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

International Nuclear Information System (INIS)

Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

2002-01-01

The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

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)

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...

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

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.)

Characteristics of Core Thermal-Hydraulic Design of SMART-P

International Nuclear Information System (INIS)

Hwang, Dae-Hyun; Seo, Kyong-Won; Kim, Tae-Wan; Lee, Chung-Chan

2006-01-01

The SMART (System-Integrated Modular Advanced ReacTor) is an integral-type advanced light water reactor which is purposed to be utilized as an energy source for sea water desalination as well as a small scale power generation. A prototype of this reactor, named SMART-P, has been studied at KAERI in order to demonstrate the relevant technologies incorporated in the SMART design. Due to the closed-channel type fuel assemblies and low mass velocity in the reactor core, the thermal hydraulic design features of SMART-P revealed fairly different characteristics in comparison with existing PWRs. The allowable operating region of the core, from the aspect of the thermal integrity of the fuel, should be primarily limited by two design parameters; critical heat flux (CHF) and fuel temperature. The occurrence of CHF may cause a sudden increase of the cladding temperature which eventually results in the fuel failure. The fuel temperature limit is relevant to a fuel failure mechanism such as a fuel centerline melting or a phase change of metallic fuels. Two phase flow instability is also an important design parameter since a flow oscillation may trigger a CHF or mechanical vibration of the channel. The characteristics of important thermal-hydraulic design parameters have been investigated for the SMART-P core with the closed-channel type fuel assemblies which contained non-square arrayed SSF (Self-sustained Square Finned) fuel rods

Cognitive Architecture of Common and Scientific Concepts

Science.gov (United States)

Tarábek, Paul

2010-07-01

The cognitive architecture of concept is a specific structure consisting of the concept core, concept periphery, the semantic frame as the meaning and the sense of the concept, and the relations among all components of this structure. The model of the cognitive architecture of scientific and common concepts is a conceptual meta-model built upon Vygotsky's concept theory, Fillmore's semantic frame, semantic triangle, on widespread ideas of the structuring of conceptual systems, and the Hestenes' Modeling Theory. The method of semantic mapping of concepts flowing from the model is designed.

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.

Legal Protection on IP Cores for System-on-Chip Designs

Science.gov (United States)

Kinoshita, Takahiko

The current semiconductor industry has shifted from vertical integrated model to horizontal specialization model in term of integrated circuit manufacturing. In this circumstance, IP cores as solutions for System-on-Chip (SoC) have become increasingly important for semiconductor business. This paper examines to what extent IP cores of SoC effectively can be protected by current intellectual property system including integrated circuit layout design law, patent law, design law, copyright law and unfair competition prevention act.

Neutronics of a mixed-flow gas-core reactor

International Nuclear Information System (INIS)

Soran, P.D.; Hansen, G.E.

1977-11-01

The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF 6 (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation

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)

Design review report for the hydrogen interlock preliminary design

International Nuclear Information System (INIS)

Corbett, J.E.

1996-01-01

This report documents the completion of a preliminary design review for the hydrogen interlock. The hydrogen interlock, a proposed addition to the Rotary Mode Core Sampling (RMCS) system portable exhauster, is 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

Core-shell designed scaffolds for drug delivery and tissue engineering.

Science.gov (United States)

Perez, Roman A; Kim, Hae-Won

2015-07-01

Scaffolds that secure and deliver therapeutic ingredients like signaling molecules and stem cells hold great promise for drug delivery and tissue engineering. Employing a core-shell design for scaffolds provides a promising solution. Some unique methods, such as co-concentric nozzle extrusion, microfluidics generation, and chemical confinement reactions, have been successful in producing core-shelled nano/microfibers and nano/microspheres. Signaling molecules and drugs, spatially allocated to the core and/or shell part, can be delivered in a controllable and sequential manner for optimal therapeutic effects. Stem cells can be loaded within the core part on-demand, safely protected from the environments, which ultimately affords ex vivo culture and in vivo tissue engineering. The encapsulated cells experience three-dimensional tissue-mimic microenvironments in which therapeutic molecules are secreted to the surrounding tissues through the semi-permeable shell. Tuning the material properties of the core and shell, changing the geometrical parameters, and shaping them into proper forms significantly influence the release behaviors of biomolecules and the fate of the cells. This topical issue highlights the immense usefulness of core-shell designs for the therapeutic actions of scaffolds in the delivery of signaling molecules and stem cells for tissue regeneration and disease treatment. Copyright © 2015 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

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…

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.

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.

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

An approach to development of structural design criteria for highly irradiated core components

International Nuclear Information System (INIS)

Nelson, D.V.

1980-01-01

The advent of the fast breeder reactor presents novel challenges in structural design and materials engineering. For instance, the core components of these reactors experience high energy neutron irradiation at elevated temperature, which causes significant time-dependent changes in material behaviour, such as a progressive loss of ductility. New structural design criteria are needed to extend elevated temperature design-by-analysis to account for these changes. Alloys best able to cope with the demands of the core operating environment are being explored and their structural behaviour characterized. The purpose of this paper is to illustrate an approach used in the development of core component structural design criteria. To do this, several design rules, plus brief rationale, from draft RDT Standards F9-7, -8 and -9 will be presented. These recently completed standards ('Structural Design Guidelines for Breeder Reactor Core Components') were prepared for the U.S. Department of Energy and represent a consensus among most organizations participating in the U.S. breeder program. (author)

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.

Bio-Inspired Multi-Functional Drug Transport Design Concept and Simulations.

Science.gov (United States)

Pidaparti, Ramana M; Cartin, Charles; Su, Guoguang

2017-04-25

In this study, we developed a microdevice concept for drug/fluidic transport taking an inspiration from supramolecular motor found in biological cells. Specifically, idealized multi-functional design geometry (nozzle/diffuser/nozzle) was developed for (i) fluidic/particle transport; (ii) particle separation; and (iii) droplet generation. Several design simulations were conducted to demonstrate the working principles of the multi-functional device. The design simulations illustrate that the proposed design concept is feasible for multi-functionality. However, further experimentation and optimization studies are needed to fully evaluate the multifunctional device concept for multiple applications.

Penn State advanced light water reactor concept

International Nuclear Information System (INIS)

Borkowski, J.A.; Smith, K.A.; Edwards, R.M.; Robinson, G.E.; Schultz, M.A.; Klevans, E.H.

1987-01-01

The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

Transient bowing of core assemblies in advanced liquid metal fast reactors

International Nuclear Information System (INIS)

Kamal, S.A.; Orechwa, Y.

1986-01-01

Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety

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.

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

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)

The passive system for reflooding of the VVER reactor core from the second-stage hydro-accumulators: design and basic design solutions

International Nuclear Information System (INIS)

Alexandr D Efanov; Sergey G Kalyakin; Andrey V Morozov; Oleg V Remizov; Vladimir M Berkovich; Victor N Krushelnitskiy; Vladimir G Peresadko; Yuri G Dragunov; Alexey K Podshibyakin; Sergey I Zaitcev

2005-01-01

Full text of publication follows: The fundamental difference in the safety assurance of the operating NPPs and those under design implies that the safety in the existing NPPs is achieved by energy-dependent (active) systems and depends on the proficiency of attending personnel. To provide safety, the new NPP designs use the physical processes proceeding in the facility without power supply; and they are unaffected by human errors. As to the safety level, the design of the new generation nuclear power plant NPP-92 relates to the class of the improved NPPs; and it applies a principle of diversity in the structure of systems responsible for critical safety functions. In accordance with the above-mentioned safety concept, the design development required a complex of experimental investigations and numerical modeling to be conducted. Among the passive safety systems of the NPP with RP-392 is the system of the second stage hydro-accumulators (GE-2). The system of the second-stage hydro-accumulators consists of four groups of hydro-accumulating tanks with a total coolant volume of 960 m 3 . The system is intended for the core flooding with coolant during 24 hours. In each group of the hydro-accumulators, the graded coolant flowrate is provided, which depends on residual heat in the reactor. The special check valves are tuned to open at the pressure drop in the circuit below 1.5 MPa. The paper presents the thermalhydraulic substantiation of the serviceability of the second-stage hydro-accumulators system for passive heat removal from the VVER reactor core and the basic design solutions on the GE-2 system. (authors)

Research on instability design method without occurring boiling transition for hyper ABWR plants of extended core power density

International Nuclear Information System (INIS)

Okamoto, T.; Hotta, A.; Ama, T.

2008-01-01

The hyper ABWR (Advanced Boiling Water Reactor) project aims to develop an advanced BWR concept that is competitive in the global market with both highly economic and safety features. Expecting plant construction within the coming ten years, a research program for substantiating the basic design of a high core power density ABWR was conducted. By inheriting the conventional ABWR design, it is possible to reduce construction costs. In order to achieve the rated core power of over 1650MWe which is almost equivalent to that of the EPR (European Pressurized Water Reactor), the core power density of ABWR will be up-rated by at least 25%. Three key subjects linked to this target were recognized. They are, (1) fuel design applicable to the high power density core, (2) improvement of the evaluation method for the coupled neutronic and thermal-hydraulic instability under a wider power-flow operating range, and (3) improvement of the steam separator performance under high quality conditions. In this paper, the second subject has been focused on. In the second subject, the uncertainty approach was introduced in the instability analysis where the best-estimate plant simulator was combined with a direct prediction of boiling transition by the sub-channel code. By employing the CSAU like method, a safety evaluation system that enables to include influences of uncertainties has been developed. Based on the correlation between the time margin for reaching the boiling transition under power oscillations and the decay ratio in the power-flow operation map, an automatic power oscillation suppressing system was designed. The set-point for activating suppression mechanisms (i.e. scram or SRI) could be determined based on this correlation. It was proposed that the present conservative acceptance criterion of the deterministic decay ratio can be replaced with a more rational one of the time margin with including uncertainties. (author)

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...

Generation-IV nuclear reactors, SFR concept

International Nuclear Information System (INIS)

Dufour, P.

2010-01-01

In this presentation author deals with development of sodium-cooled fast reactors and lead-cooled fast reactors. He concluded that: - SFR is a proved concept that has never achieved industrial deployment; - The GEN IV objectives need to reconsider the design of both the core and the reactor design : innovations are being analysed; Future design will benefit from considerable feedback of design, licensing, construction and operation of PX, SPX, etc.

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)

Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

International Nuclear Information System (INIS)

1978-09-01

The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure

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

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.

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.

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...

Embedded memory design for multi-core and systems on chip

CERN Document Server

Mohammad, Baker