Local flow distribution analysis inside the reactor pools of KALIMER-600 and PDRC performance test facility

International Nuclear Information System (INIS)

Jeong, Ji Hwan; Hwang, Seong Won; Choi, Kyeong Sik

2010-05-01

In the study, 3-dimensional thermal hydraulic analysis was carried out focusing on the thermal hydraulic behavior inside the reactor pools for both KALIMER-600 and one-fifth scale-down test facility. STAR-CD, one of the commercial CFD codes, was used to analyze 3-dimensional incompressible steady-state thermal hydraulic behavior in both designs of KALIMER-600 and the scale-down test facility. In the KALIMER-600 CFD analysis, the pressure drops in the core and IHX gave a good agreement within 1% error range. It was found that the porous media model was appropriate to analyze the pressure distribution inside reactor core and IHX. Also, a validation analysis showed the pressure drop through the porous media under the condition of 80% flow rate and thermal power was calculated 64% less than in 100% condition giving a physically reasonable analytic result. Since the temperatures in the hot-side pool and cold-side pool were estimated to be very close to 540 and 390 .deg. C specified on the design values respectively, the CFD models of heat source and sink was confirmed. Through the study, the methodology of 3-dimensional CFD analysis about KALIMER-600 has been established and proven. Performed with the methodology, the analysis data such as flow velocity, temperature and pressure distribution were compared by normalizing those data for the actual sized modeling and scale-down modeling. As a result, the characteristics of thermal hydraulic behavior were almost identical for the actual sized modeling and scale-down modeling and the similarity scaling law used in the design of the sodium test facility by KAERI was found to be correct

KALIMER design database development and operation manual

International Nuclear Information System (INIS)

Jeong, Kwan Seong; Hahn, Do Hee; Lee, Yong Bum; Chang, Won Pyo

2000-12-01

KALIMER Design Database is developed to utilize the integration management for Liquid Metal Reactor Design Technology Development using Web Applications. KALIMER Design database consists of Results Database, Inter-Office Communication (IOC), 3D CAD database, Team Cooperation System, and Reserved Documents. Results Database is a research results database for mid-term and long-term nuclear R and D. IOC is a linkage control system inter sub project to share and integrate the research results for KALIMER. 3D CAD Database is a schematic design overview for KALIMER. Team Cooperation System is to inform team member of research cooperation and meetings. Finally, KALIMER Reserved Documents is developed to manage collected data and several documents since project accomplishment

KALIMER design database development and operation manual

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kwan Seong; Hahn, Do Hee; Lee, Yong Bum; Chang, Won Pyo

2000-12-01

KALIMER Design Database is developed to utilize the integration management for Liquid Metal Reactor Design Technology Development using Web Applications. KALIMER Design database consists of Results Database, Inter-Office Communication (IOC), 3D CAD database, Team Cooperation System, and Reserved Documents. Results Database is a research results database for mid-term and long-term nuclear R and D. IOC is a linkage control system inter sub project to share and integrate the research results for KALIMER. 3D CAD Database is a schematic design overview for KALIMER. Team Cooperation System is to inform team member of research cooperation and meetings. Finally, KALIMER Reserved Documents is developed to manage collected data and several documents since project accomplishment.

Applicability of PRISM PRA Methodology to the Level II Probabilistic Safety Analysis of KALIMER-600 (I) (Core Damage Event Tree Analysis Part)

International Nuclear Information System (INIS)

Park, S. Y.; Kim, T. W.; Ha, K. S.; Lee, B. Y.

2009-03-01

The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the PSA domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of KALIMER-600 reactor. An applicability of the PSA of the PRISM plant to the KALIMER-600 has been studied. The study is confined to a core damage event tree analysis which is a part of a level 2 PSA. Assuming that the accident types, which can be developed from level 1 PSA, are same as the PRISM PRA, core damage categories are defined and core damage event trees are developed for the KALIMER-600 reactor. Fission product release fractions of the core damage categories and branch probabilities of the core damage event trees are referred from the PRISM PRA temporarily. Plant specific data will be used during the detail analysis

KALIMER preliminary conceptual design report

Energy Technology Data Exchange (ETDEWEB)

Hahn, Do Hee; Kim, Y. J.; Kim, Y. G. and others

2000-08-01

This report, which summarizes the result of preliminary conceptual design activities during Phase 1, follows the format of safety analysis report. The purpose of publishing this report is to gather all of the design information developed so far in a systematic way so that KALIMER designers have a common source of the consistent design information necessary for their future design activities. This report will be revised and updated as design changes occur and more detailed design specification is developed during Phase 2. Chapter 1 describes the KALIMER Project. Chapter 2 includes the top level design requirements of KALIMER and general plant description. Chapter 3 summarizes the design of structures, components, equipment and systems. Specific systems and safety analysis results are described in the remaining chapters. Appendix on the HCDA evaluation is attached at the end of this report.

KALIMER preliminary conceptual design report

International Nuclear Information System (INIS)

Hahn, Do Hee; Kim, Y. J.; Kim, Y. G. and others

2000-08-01

This report, which summarizes the result of preliminary conceptual design activities during Phase 1, follows the format of safety analysis report. The purpose of publishing this report is to gather all of the design information developed so far in a systematic way so that KALIMER designers have a common source of the consistent design information necessary for their future design activities. This report will be revised and updated as design changes occur and more detailed design specification is developed during Phase 2. Chapter 1 describes the KALIMER Project. Chapter 2 includes the top level design requirements of KALIMER and general plant description. Chapter 3 summarizes the design of structures, components, equipment and systems. Specific systems and safety analysis results are described in the remaining chapters. Appendix on the HCDA evaluation is attached at the end of this report

KALIMER database development (database configuration and design methodology)

International Nuclear Information System (INIS)

Jeong, Kwan Seong; Kwon, Young Min; Lee, Young Bum; Chang, Won Pyo; Hahn, Do Hee

2001-10-01

KALIMER Database is an advanced database to utilize the integration management for Liquid Metal Reactor Design Technology Development using Web Applicatins. KALIMER Design database consists of Results Database, Inter-Office Communication (IOC), and 3D CAD database, Team Cooperation system, and Reserved Documents, Results Database is a research results database during phase II for Liquid Metal Reactor Design Technology Develpment of mid-term and long-term nuclear R and D. IOC is a linkage control system inter sub project to share and integrate the research results for KALIMER. 3D CAD Database is s schematic design overview for KALIMER. Team Cooperation System is to inform team member of research cooperation and meetings. Finally, KALIMER Reserved Documents is developed to manage collected data and several documents since project accomplishment. This report describes the features of Hardware and Software and the Database Design Methodology for KALIMER

Under-Sodium Inspection Techniques for Reactor Internals of KALIMER-600 using Ultrasonic Waveguide Sensor

International Nuclear Information System (INIS)

Joo, Young Sang; Kim, Seok Hoon; Lee, Jae Han

2005-01-01

KALIMER-600 is a pool type liquid metal reactor (LMR) which is operated with a sodium coolant. The reactor internals of KALIMER-600 are submerged in a liquid sodium pool. As the liquid sodium is opaque to the light, a conventional visual inspection can not be used for observing the internal structures under a sodium condition. An under-sodium viewing (USV) technique using an ultrasonic wave should be applied for the observation of the refueling maneuver and the in-service inspection of the reactor internals. Under-sodium inspection technology utilizing ultrasonic waves has been widely developed for a visualization of the reactor core and internal components of LMR. Immersion sensors and waveguide sensors have been applied to the USV inspection. The immersion sensor has a precise imaging capability, but may have high temperature restrictions and an uncertain life. The waveguide sensor has the advantages of simplicity and reliability, but limited in its movement. The new plate-type waveguide sensor has been developed as a useful alternative to immersion sensors for USV applications. In the viewing and monitoring applications, a beam steering function of a waveguide sensor might be required. A new waveguide sensor and technique are being developed to overcome the limitations of a waveguide ultrasonic sensor. In this study, the under-sodium inspection techniques using the newly developed waveguide sensor for the reactor internal structures of KALIMER-600 is proposed

Preliminary safety analysis for key design features of KALIMER

Energy Technology Data Exchange (ETDEWEB)

Hahn, D. H.; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, S. O.; Lee, Y. B.; Jeong, K. S

2000-07-01

KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER(Korea Advanced Liquid Metal Reactor) under the long-term nuclear R and D program. In this report, descriptions of the KALIMER safety design features and safety analyses results for selected ATWS accidents are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the safety evaluation procedure for the KALIMER design is described in chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure design performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram(ATWS) have been performed to investigate the KALIMER system response to the events. They are categorized as bounding events(BEs) because of their low probability of occurrence. In chapter 4, the design of the KALIMER containment dome and the results of its performance analysis are presented. The designs of the existing LMR containment and the KALIMER containment dome have been compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core kinetics and hydraulic behavior during HCDA in chapter 5. Mathematical formulations have been developed in the framework of the modified bethe-tait method, and scoping analyses have been performed for the KALIMER core behavior during super-prompt critical excursions.

Preliminary safety design analysis of KALIMER

Energy Technology Data Exchange (ETDEWEB)

Suk, Soo Dong; Kwon, Y. M.; Kim, K. D. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-03-01

The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute(KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This document first introduces a set of safety design requirements and accident evaluation criteria established for the conceptual design of KALIMER and then summarizes some of the preliminary results of engineering and design analyses performed for the safety of KALIMER. 19 refs., 19 figs., 6 tabs. (Author)

Study on the KALIMER safety approach

International Nuclear Information System (INIS)

Kim, Eui Kwang; Han, Do Hee; Kim, Young Cheol.

1997-01-01

This study describes KALIMER's safety approach, how to establish the safety criteria and temperature limit, how to define safety evaluation events, and some safety research and development needs items. It is recommended that the KALIMER's approach to safety use seven levels of safety design and a defense-in-depth design approach with particular emphasis on inherent passive features. In order to establish as set DBEs for KALIMER safety evaluation, the procedure is explained how to define safety evaluation events. Final selection is to be determined later with the final establishment of design concepts. On the basis of preliminary studies and evaluation of the plant safety related areas, the KALIMER and PRISM have following three main difference that may require special research and development for KALIMER. (author). 7 refs., 6 tabs., 6 figs

Safety performance of preliminary KALIMER conceptual design

Energy Technology Data Exchange (ETDEWEB)

Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong [Korea atomic Energy Resarch Inst., Taejon (Korea)

1999-07-01

The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

Safety performance of preliminary KALIMER conceptual design

International Nuclear Information System (INIS)

Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong

1999-01-01

The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

Conceptual design by analysis of KALIMER seismic isolation

International Nuclear Information System (INIS)

You, Bong; Koo, Kyung Hoi; Lee, Jae Han

1996-06-01

The objectives of this report are to preliminarily evaluate the seismic isolation performance of KALIMER (Korea Advance LIquid MEtal Reactor) by seismic analyses, investigate the design feasibility, and find the critical points of KALIMER reactor structures. The work scopes performed in this study are 1) the establishment of seismic design basis, 2) the development of seismic analysis model of KALIMER, 3) the modal analysis, 4) seismic time history analysis, 5) the evaluations of seismic isolation performance and seismic design margins, and 6) the evaluation of seismic capability of KALIMER. The horizontal fundamental frequency of KALIMER reactor structure is 8 Hz, which is far remote from the seismic isolation frequency, 0.7 Hz. The vertical first and second natural frequencies are about 2 Hz and 8 Hz respectively. These vertical natural frequencies are in a dominant ground motion frequency bands, therefore these modes will result in large vertical response amplifications. From the results of seismic time history analyses, the horizontal isolation performance is great but the large vertical amplifications are occurred in reactor structures. The RV Liner has the smallest seismic design margin as 0.18. From the results of seismic design margins evaluation, the critical design change are needed in the support barrel, separation plate, and baffle plate points. The seismic capability of KALIMER is about 0.35g. This value can be increased by the design changes of the separation plate and etc.. 11 tabs., 29 figs., 7 refs. (Author) .new

Analysis of multiple failure accident scenarios for development of probabilistic safety assessment model for KALIMER-600

International Nuclear Information System (INIS)

Kim, T.W.; Suk, S.D.; Chang, W.P.; Kwon, Y.M.; Jeong, H.Y.; Lee, Y.B.; Ha, K.S.; Kim, S.J.

2009-01-01

A sodium-cooled fast reactor (SFR), KALIMER-600, is under development at KAERI. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as coolant. Its advantages are found in the aspects of an excellent uranium resource utilization, inherent safety features, and nonproliferation. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing it from the aspects of a risk informed design (RID) as well as a technology-neutral licensing (TNL). The core damage is defined as coolant voiding, fuel melting, or cladding damage. Accident scenarios which lead to the core damage should be identified for the development of a Level-1 PSA model. The SSC-K computer code is used to identify the conditions which lead to core damage. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects such as Doppler, sodium void, core axial expansion, control rod axial expansion, core radial expansion, etc. The accidents which are analyzed are the multiple failure accidents such as an unprotected transient overpower, a loss of flow, and a loss of heat sink events with degraded safety systems or functions. The safety functions to be considered here are a reactor trip, inherent reactivity feedback features, the pump coast-down, and the passive decay heat removal. (author)

DBE Analysis for KALIMER-600

International Nuclear Information System (INIS)

Ha, Kwi Seok; Jeong, Hae Young; Kwon, Young Min; Chang, Won Pyo; Lee, Yong Bum; Kim, Young II

2009-01-01

The SFR (Sodium Fast Reactor) which is being developed at KAERI (Korea Atomic Energy Research Institute) is currently divided into three types, such as, Advanced Concept 600 MWe break-even reactor and burner reactor and 1200 MWe break-even reactor. As a part of accidents analysis of the 600 MWe break-even reactor, 5 representative DBE's (Design Bases Events) are analyzed for the safety analysis. The 5 DBE's are TOP (Transient of Over Power), LOF (Loss Of Flow), LOHS (Loss Of Heat Sink), Pipe Break, and SBO (Station Black Out)

Design requirement on KALIMER control rod assembly duct

International Nuclear Information System (INIS)

Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J.

1998-03-01

This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs

Design requirement on KALIMER control rod assembly duct

Energy Technology Data Exchange (ETDEWEB)

Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J

1998-03-01

This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs.

Preliminary safety analysis for key design features of KALIMER with breakeven core

Energy Technology Data Exchange (ETDEWEB)

Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

2001-06-01

KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

Economic evaluation of KALIMER

Energy Technology Data Exchange (ETDEWEB)

Moon, Kee Hwan

1997-01-01

The main results of this study are as follows. To estimate the economic feasibility of KALIMER, the cost estimate model has been developed by using MS Excel software. Two scenarios were considered in this study. Scenario-A is composed of KALIMER options, which have FC1B (first commercial plant with 1 block), FC3B (first commercial plant with 3 blocks), NOAK1B (Nth-of-a-kind plant with 1 block), NOAK3B(Nth-of-a-kind plant with 3 blocks). The size of each block is 333 MWe. Scenario-B is comprised of PWR options, which have existing PWRs and new concepts of advanced PWR (APWR) in order to compare with KALIMER options. According to the results, the specific capital cost ($/kWe) and the levelized busbar cost (mills/kWh) for the NOAK3B option are 11% and 12% lower than that of FC3B option, respectively. These results from learning effects, scaling factors and some reductions of material and labor requirements for the NOAK3B option. And the levelized capital cost of NOAK3B option is 17%, 6% lower than that of existing PWR and APWR option, respectively. These results form shorten of construction times and labor requirements, modularization and design simplications etc. Therefore, decision and policy maker related to KALIMER development must note through the results of this study that multi-blocks design concept for its commercial plant should be considered to get the economy of scale effects. KALIMER has high competitiveness comparing to the existing PWRs and APWR. Therefore, it should be considered as a power supply option in the future in Korea. (author). 7 refs., 17 tabs., 7 figs.

Economic evaluation of KALIMER

International Nuclear Information System (INIS)

Moon, Kee Hwan.

1997-01-01

The main results of this study are as follows. To estimate the economic feasibility of KALIMER, the cost estimate model has been developed by using MS Excel software. Two scenarios were considered in this study. Scenario-A is composed of KALIMER options, which have FC1B (first commercial plant with 1 block), FC3B (first commercial plant with 3 blocks), NOAK1B (Nth-of-a-kind plant with 1 block), NOAK3B(Nth-of-a-kind plant with 3 blocks). The size of each block is 333 MWe. Scenario-B is comprised of PWR options, which have existing PWRs and new concepts of advanced PWR (APWR) in order to compare with KALIMER options. According to the results, the specific capital cost ($/kWe) and the levelized busbar cost (mills/kWh) for the NOAK3B option are 11% and 12% lower than that of FC3B option, respectively. These results from learning effects, scaling factors and some reductions of material and labor requirements for the NOAK3B option. And the levelized capital cost of NOAK3B option is 17%, 6% lower than that of existing PWR and APWR option, respectively. These results form shorten of construction times and labor requirements, modularization and design simplications etc. Therefore, decision and policy maker related to KALIMER development must note through the results of this study that multi-blocks design concept for its commercial plant should be considered to get the economy of scale effects. KALIMER has high competitiveness comparing to the existing PWRs and APWR. Therefore, it should be considered as a power supply option in the future in Korea. (author). 7 refs., 17 tabs., 7 figs

An Evaluation of the Acoustic Signal processing Techniques for Sodium-Water Reaction Detection in KALIMER-600

International Nuclear Information System (INIS)

Hur, Seop; Seong, S. H.; Kim, T. J.; Kim, S. O.; Lee, M. K.

2005-02-01

KALIMER-600 is a pool type fast breeder reactor using liquid sodium as a coolant. Although it has the several advantages such as long-term fuel cycle and enhanced safety concepts, it is possible to leak the secondary side water/steam into sodium boundary. This event could make the plant abnormal condition. One of the major design issues in KALIMER-600 is, therefore, to develop the system which can early detect the sodium-water reaction to protect the sodium-water reaction event. After evaluating the various signal processing techniques for passive acoustic leak detection, we have proposed the early leak detection logics. the signal processing techniques for evaluation were the spectral estimation using the linear modeling, the estimation error of linear modeling, the system adaptation rate using an adaptive signal processing, and the background noise cancellation using adaptive and fixed filtering. As the analysis results regarding the stationary and the cross-correlation of leak signals and background noises, the two signal systems met a wide-dense stationary process and there was only the week cross correlation relationship between two signals. It is ,therefore, possible to use the linear/harmonic modeling of signal systems, and the leak signal in sensor outputs can be discriminated. As the results of the evaluation of the various spectral estimation methods, the spectral estimation method based on autoregressive modeling was more practical comparing with other methods in the sodium-water reaction detection. The passive acoustic leak detection logics were suggested based on above evaluations. the logics consist of 3 levels; transient identification, leak determination and leak symptom identification. The simulation results using sodium-water reaction signals showed that it was possible to determine the leak at above -3dB of SNR, while between -3 dB and -10 dB of SNR the logics determined the leak symptom identification. The detection sensitivity can be enhanced

Conceptual design and assessment of in-service inspection and maintenance of KALIMER

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Kim, Seok Hun; Kim, Jong Bum; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-05-01

In the conceptual design stage of KALIMER, the philosophy and methodology of in-service inspection (ISI) and maintenance for the reactor system and components are proposed and described. The ISI and maintenance should be carried out throughout plant life to ensure the structural integrity and safety of KALIMER. The conceptual design of ISI and maintenance are performed for considering the design characteristics of KALIMER and the intents of the ASME XI Division 3. This report describes and summarizes the requirements and available methods of ISI and maintenance. The visual inspection and continuous monitoring play a great role in the in-service inspection of KALIMER. The major structures of KALIMER reactor system are designed for maintenance free operation for the plant life time and the maintenance philosophy is to replace major components rather than repair them. The assessment of the ISI accessibility and maintainability is performed and reviewed each major component. The postulated failure defects for each component are estimated and evaluated for KALIMER safety and reliability. 8 refs., 16 figs., 13 tabs. (Author)

KALIMER database development

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kwan Seong; Lee, Yong Bum; Jeong, Hae Yong; Ha, Kwi Seok

2003-03-01

KALIMER database is an advanced database to utilize the integration management for liquid metal reactor design technology development using Web applications. KALIMER design database is composed of results database, Inter-Office Communication (IOC), 3D CAD database, and reserved documents database. Results database is a research results database during all phase for liquid metal reactor design technology development of mid-term and long-term nuclear R and D. IOC is a linkage control system inter sub project to share and integrate the research results for KALIMER. 3D CAD database is a schematic overview for KALIMER design structure. And reserved documents database is developed to manage several documents and reports since project accomplishment.

KALIMER database development

International Nuclear Information System (INIS)

Jeong, Kwan Seong; Lee, Yong Bum; Jeong, Hae Yong; Ha, Kwi Seok

2003-03-01

KALIMER database is an advanced database to utilize the integration management for liquid metal reactor design technology development using Web applications. KALIMER design database is composed of results database, Inter-Office Communication (IOC), 3D CAD database, and reserved documents database. Results database is a research results database during all phase for liquid metal reactor design technology development of mid-term and long-term nuclear R and D. IOC is a linkage control system inter sub project to share and integrate the research results for KALIMER. 3D CAD database is a schematic overview for KALIMER design structure. And reserved documents database is developed to manage several documents and reports since project accomplishment

Preliminary Acceptance Criteria for Safety Analysis of KALIMER-600 SFR

International Nuclear Information System (INIS)

Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Chang, Won Pyo; Jeong, Hae Yong

2010-01-01

The KALIMER-600 event categorization in the function of occurrence frequency has been made by traditional engineering judgment with information from some reference plants such as CRBR, PRISM and EFR. The dividing line between DBE and BDBE is the frequency of 10 -7 per plant-year. Each event belongs to one of five categories based upon its nominal frequency per reactor-year (f) as a criterion. (1) Moderate frequency Event (MF): f ≥ 10 -1 (2) Infrequent Event (IE): 10 -1 > f ≥ 10 -2 (3) Unlikely Event (UE): 10 -2 > f ≥ 10 -4 (4) Extremely Unlikely Event (XU): 10 -4 > f ≥ 10 -7 (5) Beyond DBE (BDBE): > f ≥ 10 -4

Preliminary evaluation of FY98 KALIMER shielding design

Energy Technology Data Exchange (ETDEWEB)

Yoo, Jae Woon; Kang, Chang Mu; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-01-01

This report describes a preliminary evaluation of the shielding design of FY98 KALIMER. The KALIMER shielding design includes the Inner Fixed Shield of a stainless cylinder located inside the support barrel; the Radial PSDRS Shields which are three B{sub 4}C cylinders located outside the support barrel at core level; the Lower IHX shield of a cylindrical B{sub 4}C plate located above the flow guide; and Inner and Outer IHX shields of B{sub 4}C cylinders located inside and outside of the support barrel, respectively. The DORT3.1 two-dimensional transport code was used to evaluate the KALIMER shielding design. The reactor system was represented by four axial zones, each of which was modeled in the R-Z geometry. The KAFAX-F22 library was used in the analyses, which was generated from the JEF-2.2 of OECD/NEA files for LMR applications by KAERI. The performance of the KALIMER shielding design is compared against the shielding design criteria. The results indicate that the support barrel, upper grid plate, and other reactor structures meet the maximum neutron fluence and DPA limits established in the shielding design criteria. Activities of the air effluent in the PSDRS were also evaluated and are shown to satisfy the maximum permissible concentration (MPC) limits in 10 CFR Part 20. In the future, the validation of the DORT model by a detailed three dimensional calculation such as MCNP and the justification of the current shielding design limits are needed. (author). 13 refs., 23 figs., 31 tabs.

Sensitivity of Transmutation Capability to Recycling Scenarios in KALIMER-600 TRU Burner

International Nuclear Information System (INIS)

Lee, Yong Kyo; Kim, Myung Hyun

2013-01-01

The purpose of this study is to test transmutation and design feasibility of KALIMER burner caused from many limitations in recycling options; such as low recovery factors and external feed. Design impact from many recycling options will be tested as a sensitivity to various recycling process parameters under many recycling scenarios. Through this study, possibilities when Pyro-processing is realized with SFR can be expected in the recycling scenarios. For the development of sodium-cooled fast reactor(SFR) technology, prototype KALIMER plant is now under R and D stage in Korea. For the future application of SFR for waste transmutation, KALIMER core was designed for TRU burner by KAERI. Feasibility of TRU burner cannot be evaluated exactly because overall functional parameters in pyro-processing recycling process has not been verified yet. There is great possibility to accept undesirable process functions in pyro-processing. Only TRU nuclides composition a little differs between PWR SF and CANDU SF so first scenario has no problem operating SFR. In second scenario, the radiotoxicity of waste at 99% of TRU RF have to be confirmed whether it is proper level to reposit as Low and Intermediate Level Wastes or not. And the reactor safety at high RF of RE must be inspected. Not only third scenario but also several scenarios for good measure are being calculated and will be evaluated

Conceptual design of in-service inspection and maintenance for KALIMER

International Nuclear Information System (INIS)

Ju, Y. S.; Kim, S. H.; Koo, K. H.; You, B.

1999-01-01

In-service inspection and maintenance are very important for the safety and availability of nuclear power plants. The conceptual requirements of in-service inspection and maintenance should be reflected in the earlier design process for the verification of the plant operability and reliability. In this paper the fundamental approaches of the inspection and maintenance for KALIMER are established to ensure the structural integrity and operability for KALIMER. The general strategy and methodology of maintenance and inspection for the reactor system and components are proposed and described for satisfying the intents of the Section XI, Division 3, of ASME code and considering the design characteristics of KALIMER

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)

KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code

International Nuclear Information System (INIS)

Kim, Young Gyun; Kim, Young Il

2006-12-01

Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006

Preliminary conceptual design of inspection and maintenance for KALIMER reactor system

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Kim, Seok Hun; Yoo, Bong

2000-08-01

In-service inspection and maintenance are very important for improving the safety and availability of nuclear power plants. The conceptual requirements of in-service inspection and maintenance should be reflected in the earlier design process for the verification of the plant operability and reliability. In this report the fundamental approaches of the inspection and maintenance for KALIMER are established to ensure the structural integrity and operability for KALIMER. The general strategy and methodology of maintenance and inspection for the reactor system and components are proposed and described for satisfying the intents of the section XI, division 3, of ASME code and considering the design characteristics of KALIMER.

Analysis of molten fuel behavior in coolant channel during severe accidents in KALIMER

International Nuclear Information System (INIS)

Suk, Soo Dong; Lee, Yong Bum; Hahn, Do Hee

2004-11-01

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double fault initiators such as ATWS events without boiling coolant or melting fuel. For the future design of liquid metal reactor, however, the evaluation of the safety performance and the determination of containment requirements may require consideration of tripe-fault accident sequences of extremely low probability of occurrence that leads to fuel melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will required as a design requirement for the future design of LMR. For sodium-cooled core designs with metallic fuel, one of the major phenomenological modeling uncertainties to be resolved is the potential for freezing and plugging of molten metallic fuel in above- and below-core structures and possibly in inter-subassembly spaces. In this study, scoping analyses were carried out to evaluate the penetration depths in the coolant channels by molten fuel mixture during the unprotected loss-of-flow accidents in the core of the KALIMER-600. It is assumed in the analyses that a solid fuel crust would start to form upon contact with the coolant channel structure temperature of which is below the fuel solidus. The analysis results predict that the coolant channels would be plugged by the freezing molten fuel in the inlet lower shield as well as in the outlet, fission-gas-plenum region for the KALIMER-600 design

Conceptual design of data management and communication networks for KALIMER MMIS

International Nuclear Information System (INIS)

Cha, K. H.; Kwon, K. C.

1998-01-01

This paper describes the design progress for data management and communication networks to be co-operated as subsystems in KALIMER MMIS. Main functions and design bases are being established and validated for functional modules of these subsystems. Real-time data acquisition and signal validation, databases, and data logging have been designed as each functional module of data management while data interfaces of communication networks have been designed with the system information from Top-Tier Requirements for KALIMER MMIS. The conceptual design shall be refined through the iterative and detailed one

Conceptual design of data management and communication networks for KALIMER MMIS

Energy Technology Data Exchange (ETDEWEB)

Cha, K. H.; Kwon, K. C. [KAERI, Taejon (Korea, Republic of)

1998-10-01

This paper describes the design progress for data management and communication networks to be co-operated as subsystems in KALIMER MMIS. Main functions and design bases are being established and validated for functional modules of these subsystems. Real-time data acquisition and signal validation, databases, and data logging have been designed as each functional module of data management while data interfaces of communication networks have been designed with the system information from Top-Tier Requirements for KALIMER MMIS. The conceptual design shall be refined through the iterative and detailed one.

Evaluation report(1): on design criteria for KALIMER metal fuel

International Nuclear Information System (INIS)

Hwang, Woan; Lee, Byoung Oon; Kim, Young Il

2001-04-01

Fuel rods, assembly ducts and their components in KALIMER should be designed to maintain the integrities and to assure their reliable in-reactor performances under the steady state and operational transient conditions which are included in design basis category. And the fuel system must be designed with enough engineering margin to minimize and prevent the failures under ab-normal operational condition, like an accident.In this report, some design limits and the criteria for the fuel assembly ducts for KALIMER are driven by evaluating the irradiation data of metallic fuel based on experimental data from ANL in USA, CRIEPI in Japan and RIAR in Russia

Evaluation report(1): on design criteria for KALIMER metal fuel

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Lee, Byoung Oon; Kim, Young Il

2001-04-01

Fuel rods, assembly ducts and their components in KALIMER should be designed to maintain the integrities and to assure their reliable in-reactor performances under the steady state and operational transient conditions which are included in design basis category. And the fuel system must be designed with enough engineering margin to minimize and prevent the failures under ab-normal operational condition, like an accident.In this report, some design limits and the criteria for the fuel assembly ducts for KALIMER are driven by evaluating the irradiation data of metallic fuel based on experimental data from ANL in USA, CRIEPI in Japan and RIAR in Russia.

KALIMER fuel system preliminary design description

International Nuclear Information System (INIS)

Hwang, Woan; Lee, B.O.; Nam, C.; Paek, S.K.

1998-10-01

This document provides general design concepts, design basis, preliminary design specification and design technologies which are needed for designing the fuel/non-fuel rods and assembly ducts of the KALIMER fuel system. The core of LMFBR consists of driver fuel assembly, blanket assembly, reflector assembly, shielding assembly, control assembly and GEM (Gas Expansion Module) as well as USS, dummy assembly, detector assembly. These core components must be designed to withstand the high temperature, high flux for a long irradiation exposure time. Due to the high temperature and high flux, irradiation creep and swelling as well as thermal-mechanical deformation are occurred at the fuel/non-fuel system and cause the deformations of materials and the geometric deflections at fuel/non-fuel rods, assembly ducts and components. In order to overcome these intricate phenomena through the engineering design, the design basis including theoretical analysis methodologies and design considerations, material characteristics of fuel system, and the specifications and drawings of fuel/non-fuel rods and assembly ducts, respectively, are presented. This document is preliminary design description which is produced in the conceptual design stage, and does not present the detailed and finalized design data which can be for the manufacturing. (author). 22 refs

Design requirement on KALIMER blanket fuel assembly duct

International Nuclear Information System (INIS)

Hwang, Woan; Kang, H. Y.; Nam, C.; Kim, J. O.

1998-03-01

This document describes design requirements which are needed for designing the blanket fuel assembly duct of the KALIMER as design guidance. The blanket fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Blanket fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way blanket fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle is attached to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the paths for coolant inlet. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. (author). 20 refs., 4 figs

Development of MARS-LMR and Steady-state Calculation for KALIMER-600

Energy Technology Data Exchange (ETDEWEB)

Ha, K. S.; Jeong, H. Y.; Chang, W. P.; Lee, Y. B.; Jo, C. H

2007-05-15

MARS code which has been developed by coupling the RELAP and COBRA-TF in Korea Atomic Energy Research Institute has been improved in the aspects of hydraulically multi-dimensional modeling and data processing of common block using a dynamic memory allocation of FORTRAN. To use the code in the area of safety analysis of liquid metal reactor, several parts of the code have to be improved further. (1) Sodium property table including dynamic properties, such as, conductivity and viscosity, was generated to fit for the MARS code. (2) The heat transfer correlations for the liquid metal were implemented in the code. (3) The models describing the flow resistance by wire-wrap spacer in the core of LMR were applied. A MARS input data for KALIMER-600 is generated and steady-state calculation at the rated power is successfully performed. The input data can be used as a base input deck for the various transient analysis of a of PHTS, IHTS, and Tertiary system with minor revision of initial conditions and control system models.

Requirements on software lifecycle process (RSLP) for KALIMER digital computer-based MMIS design

Energy Technology Data Exchange (ETDEWEB)

Lee, Jang Soo; Kwon, Kee Choon; Kim, Jang Yeol [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-04-01

Digital Man Machine Interface System (MMIS) systems of Korea Advanced Liquid MEtal Reactor (KALIMER) may share code, data transmission, data, and process equipment to a greater degree than analog systems. Although this sharing is the basis for many of the advantages of digital systems, it also raises a key concern: a design using shared data or code has the potential to propagate a common-cause or common-mode failure via software errors, thus defeating the redundancy achieved by the hardware architectural structure. Greater sharing of process equipment among functions within a channel increases the consequences of the failure of a single hardware module and reduces the amount of diversity available within a single safety channel. The software safety plan describes the safety analysis implementation tasks that are to be carried out during the software life cycle. Documentation should exist that shows that the safety analysis activities have been successfully accomplished for each life cycle activity group. In particular, the documentation should show that the system safety requirement have been adequately addressed for each life cycle activity group, that no new hazards have been introduced, and that the software requirements, design elements, and code elements that can affect safety have been identified. Because the safety of software can be assured through both the process Verification and Validation (V and V) itself and the V and V of all the intermediate and final products during the software development lifecycle, the development of KALIMER Software Safety Framework (KSSF) must be established. As the first activity for establishing KSSF, we have developed this report, Requirement on Software Life-cycle Process (RSLP) for designing KALIMER digital MMIS. This report is organized as follows. Section I describes the background, definitions, and references of RSLP. Section II describes KALIMER safety software categorization. In Section III, we define the

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)

Seismic isolation design guidelines for KALIMER(Revision A)

International Nuclear Information System (INIS)

Yoo, B; Koo, Gyeong Hoi; Lee, J. H.

2000-04-01

The main purpose of this report is to develop the seismic isolation design guideline for KALIMER(Korea Advanced LIquid MEtal Reactor). The proposed design rules(revision A) are only applicable to the seismic isolation design with using the high damping laminated rubber bearings. When using other seismic isolation devices and applying to 3-dimensional isolation, the proposed guidelines shall be modified and added with proper research data. The rules described in this report are based on the research results performed up to now but needed to be upgraded and verified with more detail research works for the future

Preliminary conceptual design and analysis on KALIMER reactor structures

International Nuclear Information System (INIS)

Kim, Jong Bum

1996-10-01

The objectives of this study are to perform preliminary conceptual design and structural analyses for KALIMER (Korea Advanced Liquid Metal Reactor) reactor structures to assess the design feasibility and to identify detailed analysis requirements. KALIMER thermal hydraulic system analysis results and neutronic analysis results are not available at present, only-limited preliminary structural analyses have been performed with the assumptions on the thermal loads. The responses of reactor vessel and reactor internal structures were based on the temperature difference of core inlet and outlet and on engineering judgments. Thermal stresses from the assumed temperatures were calculated using ANSYS code through parametric finite element heat transfer and elastic stress analyses. While, based on the results of preliminary conceptual design and structural analyses, the ASME Code limits for the reactor structures were satisfied for the pressure boundary, the needs for inelastic analyses were indicated for evaluation of design adequacy of the support barrel and the thermal liner. To reduce thermal striping effects in the bottom are of UIS due to up-flowing sodium form reactor core, installation of Inconel-718 liner to the bottom area was proposed, and to mitigate thermal shock loads, additional stainless steel liner was also suggested. The design feasibilities of these were validated through simplified preliminary analyses. In conceptual design phase, the implementation of these results will be made for the design of the reactor structures and the reactor internal structures in conjunction with the thermal hydraulic, neutronic, and seismic analyses results. 4 tabs., 24 figs., 4 refs. (Author)

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.

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 validation report for the KALIMER core design computing system by the Monte Carlo transport theory code

International Nuclear Information System (INIS)

Lee, Ki Bog; Kim, Yeong Il; Kim, Kang Seok; Kim, Sang Ji; Kim, Young Gyun; Song, Hoon; Lee, Dong Uk; Lee, Byoung Oon; Jang, Jin Wook; Lim, Hyun Jin; Kim, Hak Sung

2004-05-01

In this report, the results of KALIMER (Korea Advanced LIquid MEtal Reactor) core design calculated by the K-CORE computing system are compared and analyzed with those of MCDEP calculation. The effective multiplication factor, flux distribution, fission power distribution and the number densities of the important nuclides effected from the depletion calculation for the R-Z model and Hex-Z model of KALIMER core are compared. It is confirmed that the results of K-CORE system compared with those of MCDEP based on the Monte Carlo transport theory method agree well within 700 pcm for the effective multiplication factor estimation and also within 2% in the driver fuel region, within 10% in the radial blanket region for the reaction rate and the fission power density. Thus, the K-CORE system for the core design of KALIMER by treating the lumped fission product and mainly important nuclides can be used as a core design tool keeping the necessary accuracy

Design study for KALIMER upper internal structure and reactor refueling system

International Nuclear Information System (INIS)

Park, Jin Ho

1996-09-01

The design study for the KALIMER upper internal structure (UIS) and reactor refueling system has been described. Two distinct features are plug-in UIS and extended refueling outage. For the UIS system, the functional, structural and material requirements have been determined and the accommodation approaches to meet these functional requirements described. For the refueling system, the functional, structural, process and I and C (Instrument and Control) requirements have been established and the accommodation approaches for the functional and process requirements described. The impact on plant availability due to extension of the refueling outage has also been investigated. The accommodation approaches for UIS system show that the design concept of the system will satisfy the functional requirements with a few design issues to be resolved, such as UIS plug in/out handling system and cask design. It is also shown that the functional and process requirements of the refueling system are achievable with the design of the IVTM cask and related transfer system and the extended refueling outage has little effect (within 1%) on the plant availability if extra refueling time do not exceed 1 week. 1 refs. (Author)

Design study for KALIMER upper internal structure and reactor refueling system

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-09-01

The design study for the KALIMER upper internal structure (UIS) and reactor refueling system has been described. Two distinct features are plug-in UIS and extended refueling outage. For the UIS system, the functional, structural and material requirements have been determined and the accommodation approaches to meet these functional requirements described. For the refueling system, the functional, structural, process and I and C (Instrument and Control) requirements have been established and the accommodation approaches for the functional and process requirements described. The impact on plant availability due to extension of the refueling outage has also been investigated. The accommodation approaches for UIS system show that the design concept of the system will satisfy the functional requirements with a few design issues to be resolved, such as UIS plug in/out handling system and cask design. It is also shown that the functional and process requirements of the refueling system are achievable with the design of the IVTM cask and related transfer system and the extended refueling outage has little effect (within 1%) on the plant availability if extra refueling time do not exceed 1 week. 1 refs. (Author).

The feasibility study on fuel types for the KALIMER

International Nuclear Information System (INIS)

Hwang, W.; Nam, C.; Yim, J. S.; Na, B. C.; Hahn, D. H.; Kim, Y. I.; Kim, Y. C.; Park, C. K.

1997-08-01

The economics of LMR is largely dependent on the construction cost of the power plant, and the fuel cycle options usually constitute 20 to 30 % of total electricity generation cost. The choice of fuel cycle technology and the fuel type is important in order to develop a LMR with better economics, performance and safety. The LMR fuel types, whose performances have been proven up to 15 at% burnup, are MOX and IFR metal fuel. The base alloy, binary (U-10% Zr) metal fuel with HT9 is used as structural materials of KALIMER. The design concept of KALIMER fuel has been established through the investigation of technical feasibilities on the fuel and recycle systems for MOX and IFR metal fuel. According to the results of comparative analysis for MOX and metal fuel, metal fuel is better than MOX in view of safety, in-reactor performance, nuclear characteristics, economics and non-proliferation, while MOX fuels have advantages in the developmental status and technical cooperation potential. The overall performance of binary (U-10% Zr) metal fuel with HT9 cladding, which is a potential start-up fuel for KALIMER, is not only superior to that of MOX fuel, but also has enough technical feasibility in its high-burnup performance, safety and economics. (author). 54 ref., 13 tabs., 20 figs

Fuel failure monitoring system design approach for KALIMER

International Nuclear Information System (INIS)

Song, Soon Ja; Hwang, I. K.; Kwon, Kee Choon

1998-01-01

Fuel Failure Monitoring System (FFMS) detects fission gas and locates failed fuels in Liquid Metal Reactor. This system comprises three subsystems; delayed neutron monitoring, cover gas monitoring, and gas tagging. The purpose of this system is to improve the integrity and availability of the liquid metal plant. In this paper, FFMS was analyzed on detection method and compared with various existing liquid metal plants. Sampling and detecting methods were classified with specific plant types. Several technologies of them was recognized and used in most liquid metal reactors. Detection technology and analysis performance, however, must be improved because of new technology when liquid metal plant is built, but the FFMS design scheme will not be changed. Thereby this paper suggests the design to implement KALIMER(Korea Advanced LIquid MEtal Reactor) FFMS

Feasibility study on the type of KALIMER coolant circulation pump

International Nuclear Information System (INIS)

Nam, H. Y.; Kim, Y. K.; Lee, Y. B.; Hwang, J. S.; Choi, S. K.

1997-07-01

The characteristics of mechanical pump and electromagnetic (EM) pump for liquid sodium coolant in a liquid metal reactor are compared and analysed as a design concept of KALIMER coolant pumps. The type of coolant circulation pump affects the selection of reactor type, economics, and reliability of reactor. Though the mechanical pump has much application experience and give satisfaction to the reliability of developed reactor type, the possibility of development is limited and its large weight and volume have a negative effect on the design of the economical liquid metal reactor. The large scale electromagnetic pump has not been verified yet, but it is expected to be demonstrated in time. Because the size of EM pump is small relative to the mechanical pump, the compact reactor design is possible. Therefore the selection of EM pump can be one of the methods to improve the economics. Since the shape of EM pump can be varied according to the arrangement of electromagnet coils, a new or unique reactor type can be developed easily in the process of KALIMER development. In the view point of economic LMR development, it is desirable to adopt the electromagnetic pump. (author). 50 refs., 11 tabs., 24 figs

Development of Fluid and I and C System Design Technology for LMR

International Nuclear Information System (INIS)

Kim, S. O.; Sim, Y. S.; Choi, S. K.

2007-06-01

The basic concept of fluid and I and C systems of KALIMER-600 was developed and the computer codes required to materialize system concept were also implemented through the R and D program. Based on the analysis results of the design characteristics for the similar reactor types developed in the foreign countries, the system design technologies with adoption of the innovative ideas were developed. With the development, expansion and reinforcement of the methodologies required according to the progress of development and design of the system and the experimental verification of the developed computer code, the excellent and innovative outcomes were produced

Development of Fluid and I and C Systems Design Technology for LMR

International Nuclear Information System (INIS)

Kim, Seong O; Sim, Y. S.; Choi, S. K.; Kim, E. K.; Wi, M. H.; Eho, J. H.; Hur, S.; Seong, S. H.; Kim, S. Y.; Jeon, W. D.

2005-03-01

The basic concept of fluid and I and C system of KALIMER-600 was developed and the computer codes required to materialize system concept were also implemented through the R and D program. Based on the analysis results of the design characteristics for the similar reactor types developed in a foreign country, the system design technologies with adoption of the innovative ideas were developed. With the development, expansion and reinforcement of the methodologies required according to the progress of development and design of the system and the experimental verification of the developed computer code, the excellent and innovative outcomes were produced

Development of fluid and I and C system design technology for LMR('03)

International Nuclear Information System (INIS)

Kim, Seong O; Sim, Yoon Sub; Choi, Seok Ki; Wi, Myung Hwan; Eoh, Jae Hyuk; Kim, Eui Kwang; Hur, Seop; Kim, Dong Hoon; Seong, Sung Hwan

2004-02-01

Based on the system design capability developed so far, our new unique reactor design concept was developed. The features are solving of a problem which existing sodium cooled reactor had, and improvement of economy and safety. Through the work, Some results were achieved, simplicity of KALIMER-600 structures, development of passive safety concept which is applicable to large sized plant, development of unique IHTS/SG combined design concept which can remove the possibility of SWR event, and optimization method of major components. Along with these results, analysis methods and computer codes, which are necessary for new design concept reactor, were developed for the self-reliance of domestic LMR technology by developing them without foreign assistance

Potential Improvements of Supercritical Recompression CO2 Brayton Cycle Coupled with KALIMER-600 by Modifying Critical Point of CO2

International Nuclear Information System (INIS)

Jeong, Woo Seok; Lee, Jeong Ik; Jeong, Yong Hoon; No, Hee Cheon

2010-01-01

Most of the existing designs of a Sodium cooled Fast Reactor (SFR) have a Rankine cycle as an electric power generation cycle. This has the risk of a sodium water reaction. To prevent any hazards from a sodium water reaction, an indirect Brayton cycle using Supercritical Carbon dioxide (S-CO 2 ) as the working fluids for a SFR is an alternative approach to improve the current SFR design. The supercritical Brayton cycle is defined as a cycle with operating conditions above the critical point and the main compressor inlet condition located slightly above the critical point of working fluid. This is because the main advantage of the cycle comes from significantly decreased compressor work just above the critical point due to high density near boundary between supercritical state and subcritical state. For this reason, the minimum temperature and pressure of cycle are just above the CO 2 critical point. In other words, the critical point acts as a limitation of the lowest operating condition of the cycle. In general, lowering the minimum temperature of a thermodynamic cycle can increase the efficiency and the minimum temperature can be decreased by shifting the critical point of CO 2 as mixed with other gases. In this paper, potential enhancement of S-CO 2 cycle coupled with KALIMER-600, which has been developed at KAERI, was investigated using a developed cycle code with a gas mixture property program

Development of KALIMER auxiliary sodium and cover gas management system

International Nuclear Information System (INIS)

Kwon, Sang Woon; Hwang, Sung Tae

1996-11-01

The objectives of this report are to develop and to describe the auxiliary liquid metal and cover gas management systems of KALIMER. the system includes following system: (1) Auxiliary liquid metal system (2) Inert gas receiving and processing system (3) Impurity monitoring and analysis system. Auxiliary liquid metal and cover gas management system of KALIMER was developed. Functions of each systems and design basis were describes. The auxiliary liquid metal system receives, transfers, and purifies all sodium used in the plant. The system furnishes the required sodium quantity at the pressure, temperature, flow rate, and purity specified by the interfacing system. The intermediated sodium processing subsystem (ISPS) provides continuous purification of IHTS sodium, as well as performs the initial fill operation for both the IHTS and reactor vessel. The primary sodium processing subsystem provides purification (cold trapping) for sodium used in the reactor vessel. The inert gas receiving and processing (IGRP) system provides liquefied and ambient gas storage, delivers inert gases of specified composition and purity at regulated flow rates and pressures to points of usage throughout the KALIMER, and accepts the contaminated gases through its vacuum facilities for storage and transfer to the gas radwaste system. Three gases are used in the KALIMER: helium, argon, and nitrogen. 11 tabs., 12 figs. (Author)

Development of KALIMER auxiliary sodium and cover gas management system

Energy Technology Data Exchange (ETDEWEB)

Kwon, Sang Woon; Hwang, Sung Tae [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-11-01

The objectives of this report are to develop and to describe the auxiliary liquid metal and cover gas management systems of KALIMER. the system includes following system: (1) Auxiliary liquid metal system (2) Inert gas receiving and processing system (3) Impurity monitoring and analysis system. Auxiliary liquid metal and cover gas management system of KALIMER was developed. Functions of each systems and design basis were describes. The auxiliary liquid metal system receives, transfers, and purifies all sodium used in the plant. The system furnishes the required sodium quantity at the pressure, temperature, flow rate, and purity specified by the interfacing system. The intermediated sodium processing subsystem (ISPS) provides continuous purification of IHTS sodium, as well as performs the initial fill operation for both the IHTS and reactor vessel. The primary sodium processing subsystem provides purification (cold trapping) for sodium used in the reactor vessel. The inert gas receiving and processing (IGRP) system provides liquefied and ambient gas storage, delivers inert gases of specified composition and purity at regulated flow rates and pressures to points of usage throughout the KALIMER, and accepts the contaminated gases through its vacuum facilities for storage and transfer to the gas radwaste system. Three gases are used in the KALIMER: helium, argon, and nitrogen. 11 tabs., 12 figs. (Author).

Review of SFR Design Safety using Preliminary Regulatory PSA Model

International Nuclear Information System (INIS)

Na, Hyun Ju; Lee, Yong Suk; Shin, Andong; Suh, Nam Duk

2013-01-01

The major objective of this research is to develop a risk model for regulatory verification of the SFR design, and thereby, make sure that the SFR design is adequate from a risk perspective. In this paper, the development result of preliminary regulatory PSA model of SFR is discussed. In this paper, development and quantification result of preliminary regulatory PSA model of SFR is discussed. It was confirmed that the importance PDRC and ADRC dampers is significant as stated in the result of KAERI PSA model. However, the importance can be changed significantly depending on assumption of CCCG and CCF factor of PDRC and ADRC dampers. SFR (sodium-cooled fast reactor) which is Gen-IV nuclear energy system, is designed to accord with the concept of stability, sustainability and proliferation resistance. KALIMER-600, which is under development in Korea, includes passive safety systems (e. g. passive reactor shutdown, passive residual heat removal, and etc.) as well as active safety systems. Risk analysis from a regulatory perspective is needed to support the regulatory body in its safety and licensing review for SFR (KALIMER-600). Safety issues should be identified in the early design phase in order to prevent the unexpected cost increase and delay of the SFR licensing schedule that may be caused otherwise

Development of Mechanical Structure Design Technology for LMR

International Nuclear Information System (INIS)

Lee, Jae Han; Joo, Young Sang; Lee, Hyeong Yeon

2007-03-01

Structural integrity and design simplifications were secured on reactor core support system, upper internal structure and core catcher of KALIMER-600. The evaluation on the suitability of high temperature and seismic design of reactor structures, and the structural integrity evaluation on reactor components and high temperature pipings are performed. The interfaces between the components and ISI accessibility are checked. Lightening of reactor building by 7%, the seismic design for 0.3g seismic loads and improvement of reactor structural design concept for KALIMER-600 have been carried out. Remote inspection technique using ultrasonic wave guide sensor was acquired as a visualization method for reactor internals under opaque sodium environments. The basic guideline on high temperature structure assessment as an assessment procedure on high temperature inelastic behaviour has been completed. In high temperature creep-fatigue test, totally 500 cycles (totally 700 hold time) were carried on cylindrical test and IHTS co-axial pipe test models. The behaviors of creep-fatigue damage and creep-fatigue crack behaviour were investigated, and the DB on the structural test were established. The seismic response tests on 19-sub assembly validation test model in air and in water were carried out, and its multi-purpose characteristics and reliability on the SAC-CORE3.0 code developed for core seismic response analysis were validated

A study on the characteristics of the decay heat removal capacity for a large thermal rated LMR design

International Nuclear Information System (INIS)

Uh, J. H.; Kim, E. K.; Kim, S. O.

2003-01-01

The design characteristics and the decay heat removal capacity according to the type of DHR (Decay Heat Removal) system in LMR are quantitatively analyzed, and the general relationship between the rated core thermal power and decay heat removal capacity is created in this study. Based on these analyses results, a feasibility of designing a larger thermal rating KALIMER plant is investigated in view of decay heat removal capacity, and DRC (Direct Reactor Cooling) type DHR system which rejects heat from the reactor pool to air is proper to satisfy the decay heat removal capacity for a large thermal rating plant above 1,000 MWth. Some defects, however, including the heat loss under normal plant operation and the lack of reliance associated with system operation should be resolved in order to adopt the total passive concept. Therefore, the new concept of DHR system for a larger thermal rating KALIMER design, named as PDRC (passive decay heat removal circuit), is established in this study. In the newly established concept of PDRC, the Na-Na heat exchanger is located above the sodium cold pool and is prevented from the direct sodium contact during normal operation. This total passive feature has the superiority in the aspect of the minimizing the normal heat loss and the increasing the operation reliance of DHR system by removing either any operator action or any external operation signal associated with system operation. From this study, it is confirmed that the new concept of PDRC is useful to the designing of a large thermal rating power plant of KALIMER-600 in view of decay heat removal capability

A study on the methodology of probabilistic safety assessment for KALIMER

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kwan Seong; Kwon, Young Min; Lee, Yong Bum; Jeong, Hae Yong; Yang, Joon Eon; Ha, Kyu Suk; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-03-01

Existing Probabilistic Safety Assessment(PSA) is a method for Light Water Reactor or Pressurized Heavy Water Reactor. Because KALIMER is different from these reactor, the new methodology of PSA need to be developed. In this paper, the PSA of Power Reactor Inherently Safety Module(PRISM) is analyzed, and Initiating Event such as Experiential Assessment, Logical Assessment and Failure Mode Effect Analysis(FMEA) is reviewed. Also, Pipe Damage Frequency Method is suggested for KALIMER. And the Reliability Physical method of Passive System, which is a chief safety system of KALIMER, is reviewed and its applicability is investigated. Finally, for the Preliminary PSA of KALIMER, Intermediate Heat Transfer System is analyzed. 23 refs., 10 figs., 13 tabs. (Author)

Development of two-dimensional hot pool model and analysis of the ULOHS accident in KALIMER design

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Bum; Jeong, K. S.; Hahn, H. D

2000-10-01

In the new version of HP2D program, the variation model of the hot pool sodium level is added so that the temperature and velocity profiles can be predicted more accurately than old version. To verify and validate the developed new version model, comparison of the MONJU experimental data with the predicted one is performed and analyzed. And also the ULOHS(Unprotected Loss of Heat Sink) accident in the KALIMER design is performed and analyzed.

Development of two-dimensional hot pool model and analysis of the ULOHS accident in KALIMER design

International Nuclear Information System (INIS)

Lee, Yong Bum; Jeong, K. S.; Hahn, H. D.

2000-10-01

In the new version of HP2D program, the variation model of the hot pool sodium level is added so that the temperature and velocity profiles can be predicted more accurately than old version. To verify and validate the developed new version model, comparison of the MONJU experimental data with the predicted one is performed and analyzed. And also the ULOHS(Unprotected Loss of Heat Sink) accident in the KALIMER design is performed and analyzed

Supercritical CO2 Brayton Cycle Energy Conversion System Coupled with SFR

International Nuclear Information System (INIS)

Cha, Jae Eun; Kim, S. O.; Seong, S. H.; Eoh, J. H.; Lee, T. H.; Choi, S. K.; Han, J. W.; Bae, S. W.

2008-12-01

This report contains the description of the S-CO 2 Brayton cycle coupled to KALIMER-600 as an alternative energy conversion system. For a system development, a computer code was developed to calculate heat balance of normal operation condition. Based on the computer code, the S-CO 2 Brayton cycle energy conversion system was constructed for the KALIMER-600. Computer codes were developed to analysis for the S-CO 2 turbomachinery. Based on the design codes, the design parameters were prepared to configure the KALIMER-600 S-CO 2 turbomachinery models. A one-dimensional analysis computer code was developed to evaluate the performance of the previous PCHE heat exchangers and a design data for the typical type PCHE was produced. In parallel with the PCHE-type heat exchanger design, an airfoil shape fin PCHE heat exchanger was newly designed. The new design concept was evaluated by three-dimensional CFD analyses. Possible control schemes for power control in the KALIMER-600 S-CO 2 Brayton cycle were investigated by using the MARS code. The MMS-LMR code was also developed to analyze the transient phenomena in a SFR with a supercritical CO 2 Brayton cycle to develop the control logic. Simple power reduction and recovery event was selected and analyzed for the transient calculation. For the evaluation of Na-CO 2 boundary failure event, a computer was developed to simulate the complex thermodynamic behaviors coupled with the chemical reaction between liquid sodium and CO 2 gas. The long term behavior of a Na-CO 2 boundary failure event and its consequences which lead to a system pressure transient were evaluated

Initial and transition cycle development for KALIMER uranium fueled core

International Nuclear Information System (INIS)

Kim, Sang Ji; Kim, Young In; Kim, Young Jin; Park, Chang Kue

1998-01-01

An economic and safe equilibrium Uranium metallic fuelled core having been established, strategic loading schemes for initial and transition cycles to early reach target equilibrium cycles are suggested for U-U and U-Pu transition cycles. An iterative method to find initial core enrichment splits is developed. With non-uniform feed enrichments at the initial core adopted, this iterative method shows KALIMER can reach Uranium equilibrium cycles just after 4 reloads, keeping feed enrichment unchanged from cycle 2. Recycling of self-generated Pu is not sufficient to make KALIMER a pure Pu equilibrium core even after 56 reloads. equilibrium cycles are suggested for U-U and U-Pu transition cycles. An iterative method to find initial core enrichment splits is developed. With non-uniform feed enrichments at the initial core adopted, this iterative method shows KALIMER can reach Uranium equilibrium cycles just after 4 reloads, keeping feed enrichment unchanged from cycle 2. Recycling of self-generated Pu is not sufficient to make KALIMER a pure Pu equilibrium core even after 56 reloads

Setup of Design Concept for the Secondary System of the Sodium Cooled Fast Reactor and Development of Computational Code for the heat balance setup

International Nuclear Information System (INIS)

Kim, E. K.; Seong, S. H.; Kim, S. O.; Eoh, J. H.; Han, J. W.; Cha, J. E.

2010-12-01

KAERI developed KALIMER-600 on it own way and now is designing the 600MWe actual sized plant for SFR. Nowadays, it is emphasizing the necessity of the evaluation for NSSS design as a part of the verification for SFR design validity. In other words, it means that should be precede the setup of the heat balance and preliminary design for SFR BOP. Turbine composition was configurated to refer SAMCHEON-PO fossil plant which have similar steam condition. The heat balance of SFR BOP was deduced to based on the NSSS boundary condition of the 600MWe actual sized plant. The algorithm of the heat balance calculation program was developed to refer preliminary heat balance data. and then, the setup of the heat balance for SFR BOP was evaluated. In the performance analysis for the preliminary heat balance of the SFR BOP, it was demonstrated that turbine characteristics are similar to reference plant, such as the SAMCHEON-PO fossil plant and the PFBR of the India

Evaluation of KALIMER IHTS piping using French RCC-MR code

International Nuclear Information System (INIS)

Lee, Hyeong Yeon; Kim, J. B.; Lee, J. H.

2001-12-01

In the present report, the evaluation of design integrity for the liquid metal reactor(LMR) of KALIMER IHTS(intermediate heat transport system) piping according to the French design guideline of RCC-MR RC3600 developed for secondary piping of LMR and the evaluation procedure was presented. The evaluation results showed that the results by the simple RC-3600 procedure of design by formula were more conservative than those of ASME section III subsection NH of the design by analysis for the class I structural components

Data management and communication networks for Man-Machine Interface System in Korea Advanced Liquid MEtal Reactor : its functionality and design requirements

International Nuclear Information System (INIS)

Cha, Kyung Ho; Park, Gun Ok; Suh, Sang Moon; Kim, Jang Yeol; Kwon, Kee Choon

1998-01-01

The DAta management and Communication NETworks(DACONET), which it is designed as a subsystem for Man-Machine Interface System of Korea Advanced LIquid MEtal Reactor(KALIMER MMIS) and advanced design concept is approached, is described. The DACONET has its roles of providing the real-time data transmission and communication paths between MMIS systems, providing the quality data for protection, monitoring and control of KALIMER and logging the static and dynamic behavioral data during KALIMER operation. The DACONET is characterized as the distributed real-time system architecture with high performance. Future direction, in which advanced technology is being continually applied to Man-Machine Interface System development and communication networks of KALIMER MMIS

Data management and communication networks for Man-Machine Interface System in Korea Advanced Liquid MEtal Reactor : its functionality and design requirements

Energy Technology Data Exchange (ETDEWEB)

Cha, Kyung Ho; Park, Gun Ok; Suh, Sang Moon; Kim, Jang Yeol; Kwon, Kee Choon [KAERI, Taejon (Korea, Republic of)

1998-05-01

The DAta management and Communication NETworks(DACONET), which it is designed as a subsystem for Man-Machine Interface System of Korea Advanced LIquid MEtal Reactor(KALIMER MMIS) and advanced design concept is approached, is described. The DACONET has its roles of providing the real-time data transmission and communication paths between MMIS systems, providing the quality data for protection, monitoring and control of KALIMER and logging the static and dynamic behavioral data during KALIMER operation. The DACONET is characterized as the distributed real-time system architecture with high performance. Future direction, in which advanced technology is being continually applied to Man-Machine Interface System development and communication networks of KALIMER MMIS.

Data management and communication networks for man-machine interface system in Korea Advanced LIquid MEtal Reactor : Its functionality and design requirements

Energy Technology Data Exchange (ETDEWEB)

Cha, Kyung Ho; Park, Gun Ok; Suh, Sang Moon; Kim, Jang Yeol; Kwon, Kee Choon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

The DAta management and COmmunication NETworks(DACONET), which it is designed as a subsystem for Man-Machine Interface System of Korea Advanced LIquid MEtal Reactor (KALIMER MMIS) and advanced design concept is approached, is described. The DACONET has its roles of providing the real-time data transmission and communication paths between MMIS systems, providing the quality data for protection, monitoring and control of KALIMER and logging the static and dynamic behavioral data during KALIMER operation. The DACONET is characterized as the distributed real-time system architecture with high performance. Future direction, in which advanced technology is being continually applied to Man-Machine Interface System development of Nuclear Power Plants, will be considered for designing data management and communication networks of KALIMER MMIS. 9 refs., 1 fig. (Author)

Data management and communication networks for man-machine interface system in Korea Advanced LIquid MEtal Reactor : Its functionality and design requirements

Energy Technology Data Exchange (ETDEWEB)

Cha, Kyung Ho; Park, Gun Ok; Suh, Sang Moon; Kim, Jang Yeol; Kwon, Kee Choon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

The DAta management and COmmunication NETworks(DACONET), which it is designed as a subsystem for Man-Machine Interface System of Korea Advanced LIquid MEtal Reactor (KALIMER MMIS) and advanced design concept is approached, is described. The DACONET has its roles of providing the real-time data transmission and communication paths between MMIS systems, providing the quality data for protection, monitoring and control of KALIMER and logging the static and dynamic behavioral data during KALIMER operation. The DACONET is characterized as the distributed real-time system architecture with high performance. Future direction, in which advanced technology is being continually applied to Man-Machine Interface System development of Nuclear Power Plants, will be considered for designing data management and communication networks of KALIMER MMIS. 9 refs., 1 fig. (Author)

Metallic fuel design development

International Nuclear Information System (INIS)

Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

1999-04-01

This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual design technology on metallic fuel

Elevated temperature design of KALIMER reactor internals accounting for creep and stress-rupture effects

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Yoo, Bong

2000-01-01

In most LMFBR (Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER (Korea Advanced Liquid Metal Reactor) reactor internal structures is carried out for normal operating conditions which have the operating temperature 530 deg. C and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME code case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects. (author)

Metallic fuel design development

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

1999-04-01

This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual

Development of subchannel analysis code MATRA-LMR for KALIMER subassembly thermal-hydraulics

International Nuclear Information System (INIS)

Won-Seok Kim; Young-Gyun Kim

2000-01-01

In the sodium cooled liquid metal reactors, the design limit are imposed on the maximum temperatures of claddings and fuel pins. Thus an accurate prediction of core coolant/fuel temperature distribution is essential to the LMR core thermal-hydraulic design. The detailed subchannel thermal-hydraulic analysis code MATRA-LMR (Multichannel Analyzer for Steady States and Transients in Rod Arrays for Liquid Metal Reactors) is being developed for KALIMER core design and analysis, based on COBRA-IV-i and MATRA. The major modifications and improvements implemented into MATRA-LMR are as follows: a) nonuniform axial noding capability, b) sodium properties calculation subprogram, c) sodium coolant heat transfer correlations, and d) most recent pressure drop correlations, such as Novendstern, Chiu-Rohsenow-Todreas and Cheng-Todreas. To assess the development status of this code, the benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR for ORNL 19-pin assembly tests and EBR-II 91-pin experiments were compared to the measurements, and to SABRE4 and SLTHEN code calculation results, respectively. In this comparison, the differences are found among the three codes because of the pressure drop and the thermal mixing modellings. Finally, the major technical results of the conceptual design for the KALIMER 98.03 core have been compared with the calculations of MATRA-LMR, SABRE4 and SLTHEN codes. (author)

The development on the methodology of the initiating event frequencies for liquid metal reactor KALIMER

International Nuclear Information System (INIS)

Jeong, K. S.; Yang, Z. A.; Ah, Y. B.; Jang, W. P.; Jeong, H. Y.; Ha, K. S.; Han, D. H.

2002-01-01

In this paper, the PSA methodology of PRISM,Light Water Reactor, Pressurized Heavy Water Reactor are analyzed and the methodology of Initiating Events for KALIMER are suggested. Also,the reliability assessment of assumptions for Pipes Corrosion Frequency is set up. The reliability assessment of Passive Safety System, one of Main Safety System of KALIMER, are discussed and analyzed

Procedures of ASME code case N-201 for KALIMER. Reactor internal structures

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Yoo, B.

2001-02-01

The main objective of this report is to describe the design procedure of ASME Boiler and Pressure Vessel Code, Code Case N-201-4, which is an elevated temperature structural design code of the Nuclear reactor internal structures, checking the criteria of stress limit, accumulated inelastic strain and deformation, creep-fatigue damage, and buckling limit. As one of examples, the creep-fatigue damage evaluations are carried out for the KALIMER reactor internal structures of baffle annulus. This report is expected to be very useful in evaluating the structural integrity of the liquid metal reactor operating under an elevated temperature

An analysis of AP600 design features

International Nuclear Information System (INIS)

Park, Jong Kyoon; Jang, Moon Heui; Hwang, Yung Dong

1994-01-01

In the aspect of engineering, passive safety system concept has improved the safety degree of nuclear power plant. Therefore, the objective of this study is to check on the possibility of the capacity upgrade of nuclear power plant in the case of adopting the passive safety system concept of AP 600. The characteristics of AP 600 are the advanced functions in ECCS, heat removal of containment building and residual heat removal under the passive safety system concept. The result of this study will become the basic data of capacity upgrade of nuclear power plant and will be widely used in second year project. (Author)

An analysis of AP600 design features

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Kyoon; Jang, Moon Heui; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); and others

1994-01-01

In the aspect of engineering, passive safety system concept has improved the safety degree of nuclear power plant. Therefore, the objective of this study is to check on the possibility of the capacity upgrade of nuclear power plant in the case of adopting the passive safety system concept of AP 600. The characteristics of AP 600 are the advanced functions in ECCS, heat removal of containment building and residual heat removal under the passive safety system concept. The result of this study will become the basic data of capacity upgrade of nuclear power plant and will be widely used in second year project. (Author).

AP600 - an ALWR conceptual design

International Nuclear Information System (INIS)

Bruce, R.A.; Vijuk, R.P.

1988-01-01

The Electric Power Research Institute is spearheading an effort to develop utility requirements for the Advanced Light Water Reactor (ALWR) plants which will become the next generation nuclear power plants for the U.S. This EPRI ALWR Program involves utilities, the U.S. Department of Energy, the U.S. Nuclear Regulatory Commission, and various industry suppliers. The ALWR Program is aimed at ALWR plants which incorporate step improvements in safety, reliability, operability and power generation costs. As part of the ALWR efforts, a Westinghouse team is conducting conceptual design development of a PWR plant design called the AP600, reflecting advanced passive safety features and the chosen 600 MWe plant output. The AP600 conceptual design provides significant improvements while employing proven component technology. This paper describes the basic reactor and primary coolant system features, the passive safety system features, and plant arrangement/construction features of AP600

Summary of Structural Concept Development and High Temperature Structural Integrity Evaluation Technology for a Gen-IV SFR

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Han; Joo, Young Sang; Lee, Hyeong Yeon (and others)

2008-04-15

The economic improvement is a hot issue as one of Gen IV nuclear plant goals. It requires many researches and development works to meet the goal by securing the same level of plant safety. One of the key research items is the increase of the plant capacity with the minimum number of components and loops. Through the successful conceptual design experience for the KALIMER-600, the structural design study for a 1200MWe large capacity of sodium-cooled fast reactor has been performed to achieve the above plant size effects. The component number and reactor structural sizing were determined based on the core and fluid system design information. Several researches were performed to reduce the construction cost of NSSS in structural point of view, for example, a simplified component arrangement, concept proposals of integrated components, a high temperature LBB application technology, and an innovative in-service inspection (ISI) tool, and a computer program development of the ASME-NH design procedure of the class 1 structure and component under high temperature over 500 .deg. C. The IHTS piping arrangement was also proposed to minimize the length through the properly locating the SG and pump by 126m. Further studies of these concepts are required to confirm on the fabricability and the structural integrity for the operating and design loads. The proposed concepts will be optimized to a unified conceptual design through several trade-off studies.

The CRDL model of SSC-K code for the safety improvement of a pool-type liquid metal-cooled reactor

International Nuclear Information System (INIS)

Jung, H. Y.; Ha, K. S.; Jang, W. P.; Hu, S.; Lee, Y. B.

2004-01-01

With the increased thermal power of KALIMER-600, it becomes important to model accurately the reactivity feedback effects due to the thermal expansion of a fuel rod and internal structure during a transient. In KALIMER design, the fuel axial expansion, core radial expansion, and the control rod drive line/reactor vessel (CRDL/RV) thermal expansion are the important reactivity feedback mechanisms. It is required to develop a more detailed CRDL/RV model for the accurate analysis of the KALIMER-600 transient because the control rod drive line of 9.5 m is immersed in the hot pool. For this a new CRDL/RV model was developed to model the effect of expansion of CRDL utilizing the temperature distribution obtained with the hot-pool 2-D model of SSC-K code. It is estimated that the developed model describes more realistically the negative reactivity insertion effect due to the initial temperature change during the UTOP transient of KALIMER-150

Westinghouse AP600 advanced nuclear plant design

International Nuclear Information System (INIS)

Gangloff, W.

1999-01-01

As part of the cooperative US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) Program and the Electric Power Research Institute (EPRI), the Westinghouse AP600 team has developed a simplified, safe, and economic 600-megawatt plant to enter into a new era of nuclear power generation. Designed to satisfy the standards set by DOE and defined in the ALWR Utility Requirements Document (URD), the Westinghouse AP600 is an elegant combination of innovative safety systems that rely on dependable natural forces and proven technologies. The Westinghouse AP600 design simplifies plant systems and significant operation, inspections, maintenance, and quality assurance requirements by greatly reducing the amount of valves, pumps, piping, HVAC ducting, and other complex components. The AP600 safety systems are predominantly passive, depending on the reliable natural forces of gravity, circulation, convection, evaporation, and condensation, instead of AC power supplies and motor-driven components. The AP600 provides a high degree of public safety and licensing certainty. It draws upon 40 years of experience in light water reactor components and technology, so no demonstration plant is required. During the AP600 design program, a comprehensive test program was carried out to verify plant components, passive safety systems components, and containment behavior. When the test program was completed at the end of 1994, the AP600 became the most thoroughly tested advanced reactor design ever reviewed by the US Nuclear Regulatory Commission (NRC). The test results confirmed the exceptional behavior of the passive systems and have been instrumental in facilitating code validations. Westinghouse received Final Design Approval from the NRC in September 1998. (author)

Assessment on the characteristics of the analysis code for KALIMER PSDRS

Energy Technology Data Exchange (ETDEWEB)

Eoh, Jae Hyuk; Sim, Yoon Sub; Kim, Seong O.; Kim, Yeon Sik; Kim, Eui Kwang; Wi, Myung Hwan [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-04-01

The PARS2 code was developed to analyze the RHR(Residual Heat Removal) system, especially PSDRS(Passive Safety Decay Heat Removal System), of KALIMER. In this report, preliminary verification and sensitivity analyses for PARS2 code were performed. From the results of the analyses, the PARS2 code has a good agreement with the experimental data of CRIEPI in the range of turbulent airside flow, and also the radiation heat transfer mode was well predicted. In this verification work, it was founded that the code calculation stopped in a very low air flowrate, and the numerical scheme related to the convergence of PARS2 code was adjusted to solve this problem. Through the sensitivity analysis on the PARS2 calculation results from the change of the input parameters, the pool-mixing coefficient related to the heat capacity of the structure in the system was improved such that the physical phenomenon can be well predicted. Also the initial conditions for the code calculation such as the hot and cold pool temperatures at the PSDRS commencing time were set up by using the transient analysis of the COMMIX code, and the surface emissivity of PSDRS was investigated and its permitted variation rage was set up. From this study, overall sensitivity characteristics of the PARS2 code were investigated and the results of the sensitivity analyses can be used in the design of the RHR system of KALIMER. 14 refs., 28 figs., 2 tabs. (Author)

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

Improved Design Concept for ensuring the Passive Decay Heat Removal Performance of an SFR

International Nuclear Information System (INIS)

Eoh, Jae Hyuk; Lee, Tae Ho; Han, Ji Woong; Kim, Seong O

2011-01-01

In order to enhance the operational reliability of a purely passive decay heat removal system in KALIMER, which is named as PDRC, three design options to prevent a sodium freezing in an intermediate decay heat removal circuit were proposed, and their feasibilities was quantitatively evaluated. For all the options, more specific design considerations were made to confirm their feasibility to properly materialize their concepts in a practical system design procedure, and the general definitions for a purely passive concept and its design features have been discussed. A numerical study to evaluate the coastdown flow effect of the primary pump was performed to figure out the early stage DHR capability inside reactor pool during a loss of normal heat sink accident. The thermal-hydraulic calculations have been made by using the COMMIX-1AR/P code, and it was found that the initiation of heat removal by DHX could be accelerated by the increase of the coastdown time but it needs a large-sized flywheel. For the demonstration of the innovative concept, a large scale sodium thermal-hydraulic test facility is currently being designed. It is very difficult to reproduce both a hydrodynamic and a thermodynamic similarity to the prototype plant if the thermal driving head is determined by structure-to-fluid heat transfer under natural circulation flow. Hence the similitude requirements for the sodium thermal-hydraulic test facility employing natural convection heat transfer were developed, and the preliminary design data of the test facility by implementing proper scaling methodologies was produced. The design restrictions imposed on the test facility and the scaling distortions of the design data to the full-scale system were also discussed

Development of a Performance Analysis Code for the Off-design conditions of a S-CO2 Brayton Cycle Energy Conversion System

International Nuclear Information System (INIS)

Yoo, Yong-Hwan; Cha, Jae-Eun; Lee, Tae-Ho; Eoh, Jae-Hyuk; Kim, Seong-O

2008-01-01

For the development of a supercritical carbon dioxide (S-CO2) Brayton cycle energy conversion system coupled to KALIMER-600, a thermal balance has been established on 100% power operating conditions including all the reactor system models such as a primary heat transport system (PHTS), an intermediate heat transport system (IHTS), and an energy conversion system. The S-CO2 Brayton cycle energy conversion system consists of a sodium-CO2 heat exchanger (Hx), turbine, high temperature recuperate (HTR), low temperature recuperate (LTR), precooler, compressor no.1, and compressor no.2. Two compressors were employed to avoid a sharp change of the physical properties near their critical point with a corresponding pressure. The component locations and their operating conditions are illustrated. Energy balance of the power conversion system in KALIMER-600 was designed with the full power condition of each component. Therefore, to predict the off-design conditions and to evaluate each component, an off-design performance analysis code should be accomplished. An off-design performance analysis could be classified into overall system control logic and local system control logic. The former means that mass flow rate and power are controlled by valves, and the latter implies that a bypass or inventory control is an admitted system balance. The ultimate goal of this study is development of the overall system control logic

AP600 design certification thermal hydraulics testing and analysis

Energy Technology Data Exchange (ETDEWEB)

Hochreiter, L.E.; Piplica, E.J.

1995-09-01

Westinghouse Electric Corporation, in conjunction with the Department of Energy and the Electric Power Research Institute, have been developing an advanced light water reactor design; the AP600. The AP600 is a 1940 Mwt, 600Mwe unit which is similar to a Westinghouse two-loop Pressurized Water Reactor. The accumulated knowledge on reactor design to reduce the capital costs, construction time, and the operational and maintenance cost of the unit once it begins to generate electrical power. The AP600 design goal is to maintain an overall cost advantage over fossil generated electrical power.

Development of Sodium Two Phase Flow Model for Kalimer Core Analysis

International Nuclear Information System (INIS)

Chang, W.P.; Hahn, Dohee

2002-01-01

An algorithm for sodium boiling is developed in order to extend the applicability of SSC-K, which is a main system analysis code for the KALIMER (Korea Advanced LIquid Metal Reactor) conceptual design. As the capability of the current SSC-K version is limited to simulation of only a single-phase sodium flow, its applicable range should not be enough to assess the fuel integrity under some of HCDA (Hypothetical Core Disruptive Accident) initiating events where sodium boiling is anticipated. The two-phase flow model similar to that used for the light water system is known to be no more effective directly to liquid metal reactors, because the phenomena observed between two reactor coolant systems are definitely different. The developing algorithm is based on a multiple-bubble slug ejection model, which allows a finite number of bubbles in a channel at any time. The present work is a continuous effort following the former study to confirm a qualitative acceptance on the model. Since the model has been applied only to the active fuel region in the former study, a part of its qualification seems to have already been demonstrated. For its application to the whole KALIMER core channel, however, the model needs to be examined the applicability to the fuel regions other than the active fuel. The present study primarily focuses on that point. In a result, although the model may be improved in a sense through the present study over the previous modeling, a clear limitation is also confirmed with the validity of the model. The further development, therefore, is required for this model to achieve its goal by resolving such limitations. (authors)

Study of allegories and proverbs used in Kalim Kashani’s Divan

Directory of Open Access Journals (Sweden)

Mohammad Mir

2016-12-01

As mentioned information, it can be found that as a famous poet in Hindi style, Kalim Kashani did not abstain using proverb and allegory in his Divan and put these literary devices in his poetries in the best way. The poet has used proverb in his Divan more than allegory and the allegory device has less appearance.

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

Directory of Open Access Journals (Sweden)

Yeong-il Kim

2013-01-01

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

The deformation analysis of the KALIMER breakeven core driver fuel pin based on the axial power profile during irradiation

International Nuclear Information System (INIS)

Lee, Dong Uk; Lee, Byoung Oon; Kim, Young Kyun; Hong, Ser Gi; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

2003-03-01

In this study, material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the End Of Life(EOL) is predicted to be 68.61% and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is 1.928%, satisfying the preliminary design criterion (3%) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc

The deformation analysis of the KALIMER breakeven core driver fuel pin based on the axial power profile during irradiation

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Uk; Lee, Byoung Oon; Kim, Young Kyun; Hong, Ser Gi; Chang, Jin Wook; Lee, Ki Bok; Kim, Young Il

2003-03-01

In this study, material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the End Of Life(EOL) is predicted to be 68.61% and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is 1.928%, satisfying the preliminary design criterion (3%) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc.

Development of the Level 1 PSA Model for PGSFR Regulatory

International Nuclear Information System (INIS)

Na, Hyun Ju; Lee, Yong Suk; Shin, Andong; Suh, Nam Duk

2014-01-01

SFR (sodium-cooled fast reactor) is Gen-IV nuclear energy system, which is designed for stability, sustainability and proliferation resistance. KALIMER-600 and PGSFR (Prototype Gen-IV SFR) are under development in Korea with enhanced passive safety concepts, e.g. passive reactor shutdown, passive residual heat removal, and etc. Risk analysis from a regulatory perspective is necessary for regulatory body to support the safety and licensing review of SFR. Safety issues should be identified in the early design phase in order to prevent the unexpected cost increase and the delay of PGSFR licensing schedule. In this respect, the preliminary PSA Model of KALIMER-600 had been developed for regulatory. In this study, the development of PSA Level 1 Model is presented. The important impact factors in the risk analysis for the PGSFR, such as Core Damage Frequency (CDF), have been identified and the related safety insights have been derived. The PSA level 1 model for PGSFR regulatory is developed and the risk analysis is conducted. Regarding CDF, LOISF frequency, uncertainty parameter for passive system CCF, loss of 125V DC control center bus and damper CCF are identified as the important factors. Sensitivity analyses show that the CDF would be differentiated (lowered) according to their values

Design-reliability assurance program application to ACP600

International Nuclear Information System (INIS)

Zhichao, Huang; Bo, Zhao

2012-01-01

ACP600 is a newly nuclear power plant technology made by CNNC in China and it is based on the Generation III NPPs design experience and general safety goals. The ACP600 Design Reliability Assurance Program (D-RAP) is implemented as an integral part of the ACP600 design process. A RAP is a formal management system which assures the collection of important characteristic information about plant performance throughout each phase of its life and directs the use of this information in the implementation of analytical and management process which are specifically designed to meet two specific objects: confirm the plant goals and cost effective improvements. In general, typical reliability assurance program have 4 broad functional elements: 1) Goals and performance criteria; 2) Management system and implementing procedures; 3) Analytical tools and investigative methods; and 4) Information management. In this paper we will use the D-RAP technical and Risk-Informed requirements, and establish the RAM and PSA model to optimize the ACP600 design. Compared with previous design process, the D-RAP is more competent for the higher design targets and requirements, enjoying more creativity through an easier implementation of technical breakthroughs. By using D-RAP, the plants goals, system goals, performance criteria and safety criteria can be easier to realize, and the design can be optimized and more rational

Colonic necrosis due to calcium polystyrene sulfonate (Kalimate not suspended in sorbitol

Directory of Open Access Journals (Sweden)

María Dolores Castillo-Cejas

2013-04-01

Full Text Available Cation-exchange resins are used in the management of hyperkalemia, particularly in patients with end-stage renal disease. These resins were associated with gastrointestinal tract lesions, especially sodium polystyrene sulfonate (Kayexalate mixed with sorbitol. We present a case of colonic necrosis after the administration of calcium polystyrene sulfonate (Kalimate not suspended in sorbitol.

Analysis of local subassembly accident in KALIMER

Energy Technology Data Exchange (ETDEWEB)

Kwon, Young Min; Jeong, Kwan Seong; Hahn, Do Hee

2000-10-01

Subassembly Accidents (S-A) in the Liquid Metal Reactor (LMR) may cause extensive clad and fuel melting and are thus regarded as a potential whole core accident initiator. The possibility of S-A occurrence must be very low frequency by the design features, and reactor must have specific instrumentation to interrupt the S-A sequences by causing a reactor shutdown. The evaluation of the relevant initiators, the event sequences which follow them, and their detection are the essence of the safety issue. Particularly, the phenomena of flow blockage caused by foreign materials and/or the debris from the failed fuel pin have been researched world-widely. The foreign strategies for dealing with the S-A and the associated safety issues with experimental and theoretical R and D results are reviewed. This report aims at obtaining information to reasonably evaluate the thermal-hydraulic effect of S-A for a wire-wrapped LMR fuel pin bundle. The mechanism of blockage formation and growth within a pin bundle and at the subassembly entrance is reviewed in the phenomenological aspect. Knowledge about the recent LMR subassembly design and operation procedure to prevent flow blockage will be reflected for KALIMER design later. The blockage analysis method including computer codes and related analytical models are reviewed. Especially SABRE4 code is discussed in detail. Preliminary analyses of flow blockage within a 271-pin driver subassembly have been performed using the SABRE4 computer code. As a result no sodium boiling occurred for the central 24-subchannel blockage as well as 6-subchannel blockage.

Supercritical Carbon Dioxide Brayton Cycle Energy Conversion System

Energy Technology Data Exchange (ETDEWEB)

Cha, Jae Eun; Kim, S. O.; Seong, S. H.; Eoh, J. H.; Lee, T. H.; Choi, S. K.; Han, J. W.; Bae, S. W

2007-12-15

This report contains the description of the S-CO{sub 2} Brayton cycle coupled to KALIMER-600 as an alternative energy conversion system. For system development, a computer code was developed to calculate heat balance of 100% power operation condition. Based on the computer code, the S-CO{sub 2} Brayton cycle energy conversion system was constructed for the KALIMER-600. Using the developed turbomachinery models, the off-design characteristics and the sensitivities of the S-CO{sub 2} turbomachinery were investigated. For the development of PCHE models, a one-dimensional analysis computer code was developed to evaluate the performance of the PCHE. Possible control schemes for power control in the KALIMER-600 S-CO{sub 2} Brayton cycle were investigated by using the MARS code. Simple power reduction and recovery event was selected and analyzed for the transient calculation. For the evaluation of Na/CO{sub 2} boundary failure event, a computer was developed to simulate the complex thermodynamic behaviors coupled with the chemical reaction between liquid sodium and CO{sub 2} gas. The long term behavior of a Na/CO{sub 2} boundary failure event and its consequences which lead to a system pressure transient were evaluated.

Supercritical Carbon Dioxide Brayton Cycle Energy Conversion System

International Nuclear Information System (INIS)

Cha, Jae Eun; Kim, S. O.; Seong, S. H.; Eoh, J. H.; Lee, T. H.; Choi, S. K.; Han, J. W.; Bae, S. W.

2007-12-01

This report contains the description of the S-CO 2 Brayton cycle coupled to KALIMER-600 as an alternative energy conversion system. For system development, a computer code was developed to calculate heat balance of 100% power operation condition. Based on the computer code, the S-CO 2 Brayton cycle energy conversion system was constructed for the KALIMER-600. Using the developed turbomachinery models, the off-design characteristics and the sensitivities of the S-CO 2 turbomachinery were investigated. For the development of PCHE models, a one-dimensional analysis computer code was developed to evaluate the performance of the PCHE. Possible control schemes for power control in the KALIMER-600 S-CO 2 Brayton cycle were investigated by using the MARS code. Simple power reduction and recovery event was selected and analyzed for the transient calculation. For the evaluation of Na/CO 2 boundary failure event, a computer was developed to simulate the complex thermodynamic behaviors coupled with the chemical reaction between liquid sodium and CO 2 gas. The long term behavior of a Na/CO 2 boundary failure event and its consequences which lead to a system pressure transient were evaluated

Marinization concept for the TRICEPT TR600 robot

Energy Technology Data Exchange (ETDEWEB)

Meyer, A.; Aust, E.; Niemann, H.R.; Santos, J.F. dos [GKSS-Forschungszentrum Geesthacht GmbH (Germany). Inst. fuer Materialforschung; Hammerin, R.; Neumann, K.E. [Neos Robotics AB, Taeby (Sweden); Gibson, D. [National Hyperbaric Centre, Aberdeen (United Kingdom)

1998-11-01

The need for automated welding repair systems of marine structures, ship hulls and nuclear installations had lead to an increasing demand for subsea robots. Considering the application of friction welding to perform underwater repairs, a TRICEPT TR600 robot has been identified as the most suitable system to withstand the high reaction forces characteristic of this process. This study reviews initially the research and development work carried out at GKSS to modify and test a Siemens-MANUTEC robot. After a description of the TRICEPT TR600 robot a marinization concept is presented and discussed in detail. Problems of galvanic corrosion in seawater are addressed in a separate chapter. The deflection of the robot in subsea water currents is estimated with a worst-case calculation. (orig.) [Deutsch] Der Wunsch, Roboter auch unter Wasser einsetzen zu koennen, waechst mit steigendem Interesse nach automatisierten Schweissverfahren fuer Reparaturen an marinen Bauwerken, Schiffsruempfen und in Kernenergieanlagen. Fuer den Einsatz von Reibschweissverfahren fuer diese Reparaturen wurde der TRICEPT TR600-Roboter ausgewaehlt, da dieser auch den charakteristisch hohen Prozesskraeften widerstehen kann. Die notwendigen Modifikationen und Pruefungen werden beispielhaft anhand des bei der GKSS modifizierten Siemens-MANUTEC-Roboters vorgestellt. Nach einer Beschreibung des TRICEPT-Roboters werden die notwendigen Umbaumassnahmen detailliert dargestellt und diskutiert. Auf die Problematik der galvanischen Korrosion in Seewasser wird in einem gesonderten Kapitel naeher eingegangen. Zusaetzlich wird eine moegliche Ablenkung des Roboters durch Wasserstroemung ueberschlaegig berechnet. (orig.)

Sodium voiding analysis in Kalimer

International Nuclear Information System (INIS)

Chang, Won-Pyo; Jeong, Kwan-Seong; Hahn, Dohee

2001-01-01

A sodium boiling model has been developed for calculations of the void reactivity feedback as well as the fuel and cladding temperatures in the KALIMER core after onset of sodium boiling. The sodium boiling in liquid metal reactors using sodium as coolant should be modeled because of phenomenon difference observed from that in light water reactor systems. The developed model is a multiple -bubble slug ejection model. It allows a finite number of bubbles in a channel at any time. Voiding is assumed to result from formation of bubbles that fill the whole cross section of the coolant channel except for liquid film left on the cladding surface. The vapor pressure, currently, is assumed to be uniform within a bubble. The present study is focused on not only demonstration of the sodium voiding behavior predicted by the developed model, but also confirmation on qualitative acceptance for the model. In results, the model catches important phenomena for sodium boiling, while further effort should be made for the complete analysis. (author)

Status of fast reactor design technology development in Korea

International Nuclear Information System (INIS)

Dohee Hahn

2000-01-01

The LMR Design Technology Development Project was approved as a national long-term R and D program in 1992 by the Korea Atomic Energy Commission (KAEC) which decided to develop and construct a LMR with the goal of developing a LMR which can serve as a long term power supplier with competitive economics and enhanced safety. Based upon the KAEC decision, the Korea Atomic Energy Research Institute (KAERI) has been developing KALIMER (Korea Advanced Liquid Metal Reactor). According to the revised National Nuclear Energy Promotion Plan of June 1997, the basic design of KALIMER will be completed by 2006 and the possibility of construction will be considered sometime during the mid 2010s. Three year Phase 1 of the LMR Design Technology Development Project was completed in March 2000 and a preliminary conceptual design report has been issued. Conceptual design of KALIMER will be developed during the Phase 2 of the Project, which will last for two years. (author)

Preliminary design of reactor coolant pump canned motor for AC600

International Nuclear Information System (INIS)

Deng Shaowen

1998-01-01

The reactor coolant pump canned motor of AC600 PWR is the kind of shielded motors with high moment of inertia, high reliability, high efficiency and nice starting performance. The author briefly presents the main feature, design criterion and technical requirements, preliminary design, computation results and analysis of performance of AC600 reactor coolant pump canned motor, and proposes some problems to be solved for study and design of AC600 reactor coolant pump canned motor

Safety design analyses of Korea Advanced Liquid Metal Reactor

International Nuclear Information System (INIS)

Suk, S.D.; Park, C.K.

2000-01-01

The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

Design of BOP Systems for the AM600

Energy Technology Data Exchange (ETDEWEB)

Ouma, Victor Otieno; Abdoelatef, M. Gomaa; Na, Hyungjooh; Field, Robert [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

Targeted for emergent nuclear countries with smaller grid capacity and/or large seasonal variation in grid frequency, the design is intended to be appropriate as an initial Nuclear Power Plant (NPP) installation. The AM600 shaftline is designed to have low capital and installation cost with robust resistance to torsional vibration and other operational insults. The proposed AM600 design for pumping condensate from the hotwell to the SGFP suction is a horizontal CP/CBP combination. This design avoids many of the problems associated with vertical condensate pumps as outlined above. Further, this design represents a robust, reliable, approach with simplified maintenance and overhauls. Finally, this design has been specified and used by a major U.S. NPP operator at twelve nuclear units with more than 10 million cumulative 'pump set' operating hours with very reliable performance. The specification of a single stage, double hung, double suction barrel pump for the SGFPs avoids design issues associated with horizontal split case pumps and with multiple impeller designs. The use of electronic VFD to drive a variable speed motor simplifies the design, fabrication, and installation of the SGFP driver. Further, it eliminates the need for a 'startup' SGFP and permits use of the SGFPs independent of having a vacuum in the main condenser. The main characteristics of the design specification for the AM600 FWHs is that they consist of a single string. Moreover, four (4) LP FWHs are located in the condenser neck. Modularizing the installation of FWHs can bring significant economies by a reduction in installation work in the field with an associated reduction in erection schedules. However, the design with a single string limits operational flexibility to take a FWH out-of-service to address tube rupture.

Increasing the reliability, availability, and maintainability of the AP600 by design

International Nuclear Information System (INIS)

Trombola, D.; Meyer, C.

1993-01-01

The AP600 design is based on providing a safe, simple, standardized, and economically competitive design with a high degree of operability and ease of maintenance. Design features such as component selection, layout, and standardization increase the probability that targeted repair times are achieved. Design requirements from the utility industry and industry design practices have established criteria for: layout, changeout and replacement of parts and components; access for major pieces of equipment; and vehicle passage. These features coupled with a solid reliability assurance and maintenance program will help the AP600 meet its objectives for operation and maintenance. The AP600 draws on the operating experience and lessons learned from the utility community through design workshops and design review interaction, as well as operating plant data from sources several sources. Internally, the AP600 program incorporates the resources of Westinghouse NSD (Nuclear Service Division), which for decades has provided refueling, steam generator, reactor coolant pump, and other operating plant services. Since the early phases of the design process, the AP600 Program has executed a comprehensive reliability, availability, and maintainability program (RAM) which dealt primarily with assessing and improving plant availability. In conjunction with this program a Probabilistic Risk Assessment (PRA) was performed and submitted to the NRC with the Standard Safety Analysis Report (SSAR) in June 1992. This paper describes how AP600 ensures that the plant has design features to enhance reliability, availability, and maintainability. The RAM program that brings the plant through the design certification phase is described

Development of fluid and I and C systems design technology

Energy Technology Data Exchange (ETDEWEB)

Sim, Yoon Sub; Park, C. K.; Kim, S. O. [and others

2000-05-01

LMR is the reactor type that makes utilization of uranium resource very efficiently and the necessity of construction of a LMR in 2020's has been raised. However, the design technology required for construction has not been secured domestically. To fulfill the necessity, study has been made for the LMR system design technology and conceptual design of KALIMER systems for fluid, instrumentation, control, and protection have been developed. Also the computer code systems for the design and analysis of the KALIMER fluid systems have been developed. These study results are to used as the starting point of the next phase LMR design technology development research.

Development of fluid and I and C systems design technology

International Nuclear Information System (INIS)

Sim, Yoon Sub; Park, C. K.; Kim, S. O.

2000-05-01

LMR is the reactor type that makes utilization of uranium resource very efficiently and the necessity of construction of a LMR in 2020's has been raised. However, the design technology required for construction has not been secured domestically. To fulfill the necessity, study has been made for the LMR system design technology and conceptual design of KALIMER systems for fluid, instrumentation, control, and protection have been developed. Also the computer code systems for the design and analysis of the KALIMER fluid systems have been developed. These study results are to used as the starting point of the next phase LMR design technology development research

Development of fluid and I and C systems design technology

Energy Technology Data Exchange (ETDEWEB)

Sim, Yoon Sub; Park, C K; Kim, S O [and others

2000-05-01

LMR is the reactor type that makes utilization of uranium resource very efficiently and the necessity of construction of a LMR in 2020's has been raised. However, the design technology required for construction has not been secured domestically. To fulfill the necessity, study has been made for the LMR system design technology and conceptual design of KALIMER systems for fluid, instrumentation, control, and protection have been developed. Also the computer code systems for the design and analysis of the KALIMER fluid systems have been developed. These study results are to used as the starting point of the next phase LMR design technology development research.

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Conceptual safety design analysis of Korea advanced liquid metal reactor

International Nuclear Information System (INIS)

Suk, S. D.; Park, C. K.

1999-01-01

The national long-term R and D program, updated in 1977, requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 Mwe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R and D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of KALIMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation. (author)

NRC confirmatory safety system testing in support of AP600 design review

International Nuclear Information System (INIS)

Rhee, G.S.; Bessette, D.E.; Shotkin, L.M.

1994-01-01

Westinghouse Electric Corporation has submitted the Advanced Passive 600 MWe (AP600) nuclear power plant design to the NRC for design certification. The Office of Nuclear Regulatory Research is proceeding to conduct confirmatory testing to help the NRC staff evaluate the AP600 safety system design. For confirmatory testing, it was determined that the cost-effective route was to modify an existing full-height, full-pressure test facility rather than build a new one. Thus, all the existing integral effects test facilities, both in the US and abroad, were screened to select the best candidate. As a result, the ROSA-V (Rig of Safety Assessment-V) test facility located in the Japan Atomic Energy Research Institute (JAERI) was chosen. However, because of some differences in design between the existing ROSA-V facility and the AP600, the ROSA-V is being modified to conform to the AP600 safety system design. The modification work will be completed by the end of this year. A series of facility characterization tests will then be performed in January 1994 for the modified part of the facility before the main test series is initiated in February 1994. A total of 12 tests will be performed in 1994 under Phase I of this cooperative program with JAERI. Phase II testing is being considered to be conducted in 1995 mainly for beyond-design-basis accident evaluation

Comparison of Design Concepts for SFR under Development

Energy Technology Data Exchange (ETDEWEB)

Suh, Namduk; Choi, Yongwon; Bae, Moohoon; Shin, Andong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

The goal of ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) with a capacity of 600 MWe is to study the technical demonstration that can be scaled up to commercial reactor. It was expected that the success of ASTRID project could eventually lead to operation of industrial reactor around 2040. On 2012, ASTRID designer has submitted the DOrS (Dossier d’Orientations de Sûreté, Safety Orientation Document) for ASTRID to IRSN and IRSN has issued a report after reviewing the DOrS. The report DOrS itself is not available publicly, intellectual property might be the reason, but the review document of IRSN is open to public, so we can understand the basic concept of ASTRID by IRSN report. The DOrS of ASTRID and the TTR for PGSFR have not the same format and also the same purpose, so it is not easy to compare the two design concepts directly. But, still, we think the concepts could be compared in a very general way. Thus, in this paper we have presented the very short comparison results of the two SFR design. Our opinion after first reviewing the TTR is that the PGSFR needs to be designed in a more systematic way. The requirements are coming basically from the previous document used for SMART licensing and do not show prototype reactor specific characters.

Design criteria for the electrical system in advanced passive reactors. Special features of the AP-600 Reactor

International Nuclear Information System (INIS)

Moraleda Lopez, A.

1997-01-01

The design of the electrical system of an Passive Advanced Reactor is determined by the concept of passive actuation of safety systems, simplification of process systems and optimisation of equipment performance. The system that results from these criteria is very different to those designed for present plants. The main differences are: No class 1E alternating current systems No emergency diesel generators Fewer safety and non-safety class electricity consumers System for continuous monitoring of battery status Use of electronic speed regulators for reactor feedwater pump motors Outsite battery backup safety power supply Motor-operated valves are the only safety electrical actuators Portable power supply for post 72 hour equipment This paper develops these concepts and applies them to the AP-600 project and describes the electrical system of this type of plant. (Author)

A comparative neutronic analysis of KALIMER breeder core using Na or Pb-Bi coolant

International Nuclear Information System (INIS)

Yoo, J. W.; Kim, S. J.; Kim, Y. I.

2000-01-01

A comparative neutronic study has been conducted on KALIMER breeder core according to the replacement of sodium coolant by Pb-Bi coolant. Since the atomic weight of Pb and Bi is about 9 times heavier than that of Na, the energy loss by neutron colliding with Pb-Bi nucleus will be very small. Therefore, the reactor with Pb-Bi coolant will have a harder neutron spectrum than that with Na coolant. Consequently, the breeding ratio and burnup reactivity swing is expected to be enhanced. In addition, when Pb-Bi coolant is voided, a negative coolant void coefficient can be obtained by the net effects of smaller spectrum hardening and large neutron leakage. As a result, the breeding ratio was increased from 1.18 to 1.23 and burnup reactivity swing was reduced from 631 pcm to 150 pcm. When the coolant in the whole region of active core is voided, the coolant void coefficient was found to be -539 and -264 pcm at BOEC and EOEC, respectively. In the local voided case, the smaller coolant void coefficient was obtained than that of Na coolant. Accordingly, the use of Pb-Bi coolant in KALIMER gives an advantage of higher breeding ratio, smaller burnup reactivity swing and negative coolant void coefficient without any significant degradation of nuclear performance

Condenser Design for the Proposed AM600 NPP

Energy Technology Data Exchange (ETDEWEB)

Rahman, Md. Mizanur; Abdallah, Khaled Atya Ahmed; Field, Robert M. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

The design goals are to make the condenser more robust and compact with a reduced component count. The AM600 condenser design also has new features as described below. Considering that the minimum heat sink temperature for potentially emergent nuclear countries is on the order of 21.deg. C or higher, a turbine design with a single low pressure rotor can be considered without sacrificing thermal efficiency. The condenser back pressure range for the considered markets is on the order of 2 to 3 in-HgA. With these boundary conditions, the AM600 condenser duty can be met with a single pressure zone design with a total of eight (8) titanium tube bundles (four (4) per pass) divided into four isolable sections. Due to the compact design (i.e., accepting exhaust from only one low pressure cylinder), both axial ends of the condenser are unobstructed and available for attachment of extended flash chambers, diverting inflows away from the tube bundles. The single shell design of this condenser then allows for an innovative design feature, namely the extended flash chambers. This permits the routing of dump, drain, vent, and bypass flows directly to these chambers, bypassing the condenser shell. Within the condenser shell, this design eliminates impingement plates, impingement boxes, and spargers. Failure of these components represents an ongoing source of condenser tube damage in operating nuclear units, requiring significant resources for outage inspections. The extended flash chamber approach also has a number of other advantages as delineated above.

Improvements in nuclear plant staffing resulting from the AP600 design programme

International Nuclear Information System (INIS)

Mycoff, C.

2001-01-01

The staffing for a single-unit AP600 is estimated to require a staff for operation and maintenance about 32% smaller than current generation power plants of similar size. These staffing reductions are driven primarily by various features incorporated into the AP600 plant design. (author)

Design of the AM600 Turbine-Generator for NPPs in Emerging Markets

Energy Technology Data Exchange (ETDEWEB)

Alexandru, Bogdan; Abdoelatef, M. Gomaa; Field, Robert [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

In this paper, preliminary analysis related to: (i) the T/G steam flow path, and (ii) the turbine cycle heat balance is examined. Analysis of global electric markets indicates that the current and near term capacity of electrical grids for many developing countries (e.g., Bangladesh, Kenya, Vietnam, Malaysia) is insufficient to reliably incorporate Nuclear Power Plants (NPPs) with large unit sizes (e.g., >1000 MWe). Thus a modern NPP design with a smaller output (-600 MWe) is of interest. To address conditions for such markets, the 'AM600' Turbine-Generator (T/G) design proposed here represents a 600 MWe design which is robust and supports a simplified steam cycle. The proposed shaftline starts with a determination of the number of flows, followed by a determination of the number of high and low pressure stages, followed by heat balance analysis. The conceptual design for the AM600 T/G offers the following: • a stiff shaftline which can offer robust performance in smaller grids lacking optimal stability relative to grid disturbances and frequency variation, • a simplified approach to T/G fabrication, installation, operation, testing, inspections, and maintenance due to design with a single LPT cylinder while maintaining high thermal efficiency, and • a reduced component count for MSRs, FWHs, and power train pumps and drivers (and associated support system components) resulting in lower capital outlays, simplified operations, and further reducing the maintenance, testing, and inspection burden.

Design of the AM600 Turbine-Generator for NPPs in Emerging Markets

International Nuclear Information System (INIS)

Alexandru, Bogdan; Abdoelatef, M. Gomaa; Field, Robert

2015-01-01

In this paper, preliminary analysis related to: (i) the T/G steam flow path, and (ii) the turbine cycle heat balance is examined. Analysis of global electric markets indicates that the current and near term capacity of electrical grids for many developing countries (e.g., Bangladesh, Kenya, Vietnam, Malaysia) is insufficient to reliably incorporate Nuclear Power Plants (NPPs) with large unit sizes (e.g., >1000 MWe). Thus a modern NPP design with a smaller output (-600 MWe) is of interest. To address conditions for such markets, the 'AM600' Turbine-Generator (T/G) design proposed here represents a 600 MWe design which is robust and supports a simplified steam cycle. The proposed shaftline starts with a determination of the number of flows, followed by a determination of the number of high and low pressure stages, followed by heat balance analysis. The conceptual design for the AM600 T/G offers the following: • a stiff shaftline which can offer robust performance in smaller grids lacking optimal stability relative to grid disturbances and frequency variation, • a simplified approach to T/G fabrication, installation, operation, testing, inspections, and maintenance due to design with a single LPT cylinder while maintaining high thermal efficiency, and • a reduced component count for MSRs, FWHs, and power train pumps and drivers (and associated support system components) resulting in lower capital outlays, simplified operations, and further reducing the maintenance, testing, and inspection burden

Initial performance assessment of the Westinghouse AP600 containment design and related safety issues

International Nuclear Information System (INIS)

Nicolette, V.F.; Washington, K.E.; Tills, J.L.

1991-01-01

This work summarizes the Westinghouse AP600 advanced reactor design assessment calculations performed to date with the CONTAIN code. Correlations for modeling the important heat transfer phenomena are discussed as well. A CONTAIN model of the AP600 was constructed for design basis accident (DBA) calculations. Insights gained from modeling of the smaller-scale Westinghouse Integral Test Facility were incorporated in the development of the AP600 model. The results of the DBA calculations are compared to the results of other researchers to serve as a point of reference for future severe accident calculations. The CONTAIN calculations are reviewed to examine several parameters/phenomena of interest. The results of the calculations are also used to identify limitations of the CONTAIN code regarding application to advanced reactor containment designs. The most recent heat transfer correlations available in the literature are assessed for use in the flow regimes and geometries applicable to the AP600. Use of one of these correlations in CONTAIN may allow for a more accurate assessment of the AP600

Designing Pu600 for Authentication

International Nuclear Information System (INIS)

White, G.

2008-01-01

Many recent Non-proliferation and Arms Control software projects include an authentication component. Demonstrating assurance that software and hardware performs as expected without hidden 'back-doors' is crucial to a project's success. In this context, 'authentication' is defined as determining that the system performs only its intended purpose and performs that purpose correctly and reliably over many years. Pu600 is a mature software solution for determining the presence of Pu and the ratio of Pu240 to Pu239 by analyzing the gamma ray spectra in the 600 KeV region. The project's goals are to explore hardware and software technologies which can by applied to Pu600 which ease the authentication of a complete, end-to-end solution. We will discuss alternatives and give the current status of our work

Design of the MYRRHA 17-600 MeV Superconducting Linac

CERN Document Server

Biarrotte, J-L; Bouly, F; Carneiro, J-P; Vandeplassche, D

2013-01-01

The goal of the MYRRHA project is to demonstrate the technical feasibility of transmutation in a 100MWth Accelerator Driven System (ADS) by building a new flexible irradiation complex in Mol (Belgium). The MYRRHA facility requires a 600 MeV accelerator delivering a maximum proton flux of 4 mA in continuous operation, with an additional requirement for exceptional reliability. This paper will briefly describe the beam dynamics design of the main superconducting linac section which covers the 17 to 600 MeV energy range and requires enhanced fault-tolerance capabilities.

Designing Pu600 for Authentication

Energy Technology Data Exchange (ETDEWEB)

White, G

2008-07-10

Many recent Non-proliferation and Arms Control software projects include an authentication component. Demonstrating assurance that software and hardware performs as expected without hidden 'back-doors' is crucial to a project's success. In this context, 'authentication' is defined as determining that the system performs only its intended purpose and performs that purpose correctly and reliably over many years. Pu600 is a mature software solution for determining the presence of Pu and the ratio of Pu240 to Pu239 by analyzing the gamma ray spectra in the 600 KeV region. The project's goals are to explore hardware and software technologies which can by applied to Pu600 which ease the authentication of a complete, end-to-end solution. We will discuss alternatives and give the current status of our work.

Seismic response time history analyses for KALIMER building with a horizontal and vertical seismic isolation

International Nuclear Information System (INIS)

Lee, J. H.; Yoo, B.; Koo, K. H.

2001-01-01

The seismic response time history analyses for the lumped mass models of KALIMER reactor building with a horizontal and vertical seismic isolation are performed for Artificial Time History and Kobe earthquake. The vertical amplification by the horizontal isolation is reduced by a vertical isolation for both earthquakes. The 3% viscous damping and the vertical isolation frequency of 1.5Hz gives a reduced vertical response compared to the fixed base condition at reactor support, and the 9% viscous damping to Kobe earthquake is required to get an equivalent vertical response with a fixed base condition

Seismic response time history analyses for KALIMER building with a horizontal and vertical seismic isolation

Energy Technology Data Exchange (ETDEWEB)

Lee, J. H.; Yoo, B.; Koo, K. H. [KAERI, Taejon (Korea, Republic of)

2001-05-01

The seismic response time history analyses for the lumped mass models of KALIMER reactor building with a horizontal and vertical seismic isolation are performed for Artificial Time History and Kobe earthquake. The vertical amplification by the horizontal isolation is reduced by a vertical isolation for both earthquakes. The 3% viscous damping and the vertical isolation frequency of 1.5Hz gives a reduced vertical response compared to the fixed base condition at reactor support, and the 9% viscous damping to Kobe earthquake is required to get an equivalent vertical response with a fixed base condition.

Development Perspective of Regulatory Audit Code System for SFR Nuclear Safety Evaluation

Energy Technology Data Exchange (ETDEWEB)

Bae, Moo Hoon; Lee, Gil Soo; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2012-05-15

A sodium-cooled fast reactor (SFR) in Korea is based on the KALIMER-600 concept developed by KAERI. Based on 'Long-term R and D Plan for Future Reactor Systems' which was approved by the Korea Atomic Energy Commission in 2008, the KAERI designer is scheduled to apply the design certification of the prototype SFR in 2017. In order to establish regulatory infrastructure for the licensing of a prototype SFR, KINS has develop the regulatory requirements for the demonstration SFR since 2010, and are scheduled to develop the regulatory audit code systems in regard to core, fuel, and system, etc. since 2012. In this study, the domestic code systems used for core design and safety evaluation of PWRs and the nuclear physics and code system for SFRs were briefly reviewed, and the development perspective of regulatory audit code system for SFR nuclear safety evaluation were derived

Design of Passive Decay Heat Removal System using Mercury Thermosyphon for SFR

Energy Technology Data Exchange (ETDEWEB)

You, Byung Hyun; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2013-10-15

In this study, thermosyphon application is suggested to accomplish the fully passive safety grade system and compactness of components via enhance the heat removal performance. A two-phase evaporating thermosyphon operates when the evaporator is heated, the working fluid start boiling, the vapor that is formed moves to the condenser, where it is condensed on the walls, giving up the heat of phase change to the cooling fluid. Gravity forces cause the condensate to condensed liquid flow to the evaporator again. These processes occur continuously, which causes transfer of heat from evaporator to condenser vice versa. After the thermal design and performance evaluation, the results were compared with the performance of conventional DRACS system. For the same amount of decay heat removal performance of PDRC system of KALIMER-600 mercury thermosyphon system can archive around 30∼50% of compactness. For the detailed design, improved analytical model and experimental data for the validation will be required to specify the new DHR system.

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)

Advanced passive PWR AC-600: Development orientation of nuclear power reactors in China for the next century

International Nuclear Information System (INIS)

Huang Xueqing; Zhang Senru

1999-01-01

Based on Qinshan II Nuclear Power Plant that is designed and constructed by way of self-reliance, China has developed advanced passive PWR AC-600. The design concept of AC-600 not only takes the real situation of China into consideration, but also follows the developing trend of nuclear power in the world. The design of AC-600 has the following technical characteristics: Advanced reactor: 18-24 month fuel cycle, low neutron leakage, low power density of the core, no any penetration in the RPV below the level of the reactor coolant nozzles; Passive safety systems: passive emergency residual heat removal system, passive-active safety injection system, passive containment cooling system and main control room habitability system; System simplified and the number of components reduced; Digital I and C; Modular construction. AC-600 inherits the proven technology China has mastered and used in Qirtshan 11, and absorbs advanced international design concepts, but it also has a distinctive characteristic of bringing forth new ideas independently. It is suited to Chinese conditions and therefore is expected to become an orientation of nuclear power development by self-reliance in China for the next century. (author)

Development of self-actuated shutdown system using curie point electromagnet

International Nuclear Information System (INIS)

Kim, Tae Ryong; Park, Jin Ho

1999-01-01

An innovative concept for a passive reactor shutdown system, so called self-actuated shutdown system (SASS), is inevitably required for the inherent safety in liquid metal reactor, which is designed with the totally different concept from the usual reactor shutdown system in LWR. SASS using Curie point electromagnet (CPEM) was selected as the passive reactor shutdown system for KALIMER (Korea Advanced Liquid Metal Reactor). A mock-up of the SASS was designed, fabricated and tested. From the test it was confirmed that the mockup was self-actuated at the Curie point of the temperature sensing material used in the mockup. An articulated control rod was also fabricated and assembled with the CPEM to confirm that the control rod can be inserted into core even when the control rod guide tube is deformed due to earthquake. The operability of SASS in the actual sodium environment should be confirmed in the future. All the design and test data will be applied to the KALIMER design. (author)

Studies for the layout and technical conception of a two-circuit HTR power plant of 600 MWsub(el) under public utilizer aspects

International Nuclear Information System (INIS)

Schuetten, R.

1981-01-01

In this study concerning conceptions for a nuclear power plant of 600 MWsub(el) with high-temperature reactor a conception for a HTR-nuclear power plant of 600 MWsub(el) to be built in the Federal Republic of Germany in future is developed on the basis of operating experience with the 15-MW-AVR-experimental nuclear power plant, the construction of the THTR-300 nuclear power plant and the gas-cooled reactors of English, French and American origin. This report gives a survey of the most important findings and the requirements on behalf of the public utilities for a nuclear power plant with high-temperature reactor with the dimensions of 600 MWsub(el). The examination of the utilities basic requirements for a power plant and the experience made during the licensing procedure led to this technical and safety conception for a HTR nuclear power plant with spherical fuel elements. In addition, the questions of the possibility of recurrent tests and of repairing safety components and also the future shut-down of the power plant, which are important for the public utilities, are examined. (orig./GL) [de

600 kV modulator design for the SLAC Next Linear Collider Test Accelerator

International Nuclear Information System (INIS)

Harris, K.; de Lamare, J.; Nesterov, V.; Cassel, R.

1992-07-01

Preliminary design for the SLAC Next Linear Collider Test Accelerator (NLCTA) requires a pulse power source to produce a 600 kV, 600 A, 1.4 μs, 0.1% flat top pulse with rise and fall times of approximately 100 ns to power an X-Band klystron with a microperveance of 1.25 at ∼ 100 MW peak RF power. The design goals for the modulator, including those previously listed, are peak modulator pulse power of 340 MW operating at 120 Hz. A three-stage darlington pulse-forming network, which produces a >100 kV, 1.4 μs pulse, is coupled to the klystron load through a 6:1 pulse transformer. Careful consideration of the transformer leakage inductance, klystron capacitance, system layout, and component choice is necessary to produce the very fast rise and fall times at 600 kV operating continuously at 120 Hz

Measurement and flow visualization research of thermal hydraulic characteristics for the SFR reactor Vessel

International Nuclear Information System (INIS)

Cha, J. E.; Kim, S. O.; Choi, H. L.; Kim, H. B.; Kim, H. W.; Lee, S. H.

2012-01-01

In this report, the thermal hydraulic and flow visualization experiment was described for the KALIMER-600 water-scaled model. In order to investigate a thermal hydraulic characteristics for the SFR KALIMER-600, which has been conceptually designed in the KAERI, a water-scaled 1/10 reactor vessel model was designed and prepared through the scaling analysis during three-years research. In this research, SFR Photos system, which has inherently very complicated the internal structures, was fabricated with a transparent vessel. It was shown that a serious of thermal hydraulic test was conducted within a short period if modeled with water than sodium. Natural circulation test was successfully performed with the modeled heater assembly and heat exchanger system coupled with cooling system. The water-scaled RSV experimental facility made in this research could be used to study the USA development for the future SFR system and utilized to analyze the flow characteristics before changing a main internal part of Photos system. It could also be used to test a pool-inspection study and a sensor selection study before large scale sodium experiment. The PCV system prepared in this research could be utilized to test other TSH experiment and temperature field measurement

Multi dimensional analysis of Design Basis Events using MARS-LMR

International Nuclear Information System (INIS)

Woo, Seung Min; Chang, Soon Heung

2012-01-01

Highlights: ► The one dimensional analyzed sodium hot pool is modified to a three dimensional node system, because the one dimensional analysis cannot represent the phenomena of the inside pool of a big size pool with many compositions. ► The results of the multi-dimensional analysis compared with the one dimensional analysis results in normal operation, TOP (Transient of Over Power), LOF (Loss of Flow), and LOHS (Loss of Heat Sink) conditions. ► The difference of the sodium flow pattern due to structure effect in the hot pool and mass flow rates in the core lead the different sodium temperature and temperature history under transient condition. - Abstract: KALIMER-600 (Korea Advanced Liquid Metal Reactor), which is a pool type SFR (Sodium-cooled Fast Reactor), was developed by KAERI (Korea Atomic Energy Research Institute). DBE (Design Basis Events) for KALIMER-600 has been analyzed in the one dimension. In this study, the one dimensional analyzed sodium hot pool is modified to a three dimensional node system, because the one dimensional analysis cannot represent the phenomena of the inside pool of a big size pool with many compositions, such as UIS (Upper Internal Structure), IHX (Intermediate Heat eXchanger), DHX (Decay Heat eXchanger), and pump. The results of the multi-dimensional analysis compared with the one dimensional analysis results in normal operation, TOP (Transient of Over Power), LOF (Loss of Flow), and LOHS (Loss of Heat Sink) conditions. First, the results in normal operation condition show the good agreement between the one and multi-dimensional analysis. However, according to the sodium temperatures of the core inlet, outlet, the fuel central line, cladding and PDRC (Passive Decay heat Removal Circuit), the temperatures of the one dimensional analysis are generally higher than the multi-dimensional analysis in conditions except the normal operation state, and the PDRC operation time in the one dimensional analysis is generally longer than

Design of a Mars Airplane Propulsion System for the Aerial Regional-Scale Environmental Survey (ARES) Mission Concept

Science.gov (United States)

Kuhl, Christopher A.

2008-01-01

The Aerial Regional-Scale Environmental Survey (ARES) is a Mars exploration mission concept that utilizes a rocket propelled airplane to take scientific measurements of atmospheric, surface, and subsurface phenomena. The liquid rocket propulsion system design has matured through several design cycles and trade studies since the inception of the ARES concept in 2002. This paper describes the process of selecting a bipropellant system over other propulsion system options, and provides details on the rocket system design, thrusters, propellant tank and PMD design, propellant isolation, and flow control hardware. The paper also summarizes computer model results of thruster plume interactions and simulated flight performance. The airplane has a 6.25 m wingspan with a total wet mass of 185 kg and has to ability to fly over 600 km through the atmosphere of Mars with 45 kg of MMH / MON3 propellant.

Workshop UNK-600 (proceedings); Materialy rabochego soveshchaniya UNK-600

Energy Technology Data Exchange (ETDEWEB)

Zajtsev, A M; Bitykov, S I [eds.

1994-12-31

Proceedings are presented of the workshop devoted to the accelerating storage complex of IHEP (UNK-600). In the first section is given the information on the present status of the UNK-600 and particle channels design and on the adopted experiment NEPTUN-A. In the papers of the second section are discussed hadron physics investigations at 600 GeV. Experiments in the neutrino and muon beams are analyzed. A possible program of studying the charged kaon rare decays is described.

The development of technologies of safety analysis for LMR ('03)

Energy Technology Data Exchange (ETDEWEB)

Lee, Y. B.; Suk, S. D.; Chang, W. P.; Kwon, Y. M.; Jeong, H. Y.; Ha, K. W.; Heo, S

2004-03-01

The developmental objectives of the project, 'The development of safety analysis techniques in LMR', are the code development for the subchannel blockage analysis, the code development for the system transient analysis, the code development for the HCDA(Hypothetical Core Disruptive Accident) analysis, the preliminary safety analysis for KALIMER-600 equipped with the components of new concepts, and the establishment of data base. The purpose of the analysis for subchannel blockage in the subassembly of LMR is to represent quantitatively that the maximum damage due to the accident is within the safety criteria. The computational program should be developed to simulate the thermal hydraulic phenomena and to verify the safety of LMR for the accident. For the purpose, the hybrid scheme has been implemented into the MATRA-LMR code based on the upwind scheme to analyze the various flow fields occurred in the subchannel blockage accident. The turbulent mixing models using the CFX code were assessed to compute more precisely the heat transfer between subchannels. Through this assessment, empirical correction factors of 1.7 for the heat conduction, 0.006 for the turbulent mixing coefficient were obtained. The distributed resistance model instead of wire forcing function has been developed to represent the more exact flow field due to wire-wrap. Other models, such as heat conductor model and various turbulent mixing model, have been implemented into the MATRA-LMR. The ORNL THORS 19-Pin FFM-5B tests have been assessed to validate above new models using the improved MATRA-LMR. The results using MATRA-LMR were well agreed with the experimental data. The subchannel blockage accidents which assumed to be occurred at the three locations for the conceptual plant of KALIMER-600 have been analysed according to blockage size using the MATRA-LMR code. The results of calculations for the design basis events which 6 subchannels were blocked showed the margins of the 290 7.dog. C

The development of technologies of safety analysis for LMR ('03)

International Nuclear Information System (INIS)

Lee, Y. B.; Suk, S. D.; Chang, W. P.; Kwon, Y. M.; Jeong, H. Y.; Ha, K. W.; Heo, S.

2004-03-01

The developmental objectives of the project, 'The development of safety analysis techniques in LMR', are the code development for the subchannel blockage analysis, the code development for the system transient analysis, the code development for the HCDA(Hypothetical Core Disruptive Accident) analysis, the preliminary safety analysis for KALIMER-600 equipped with the components of new concepts, and the establishment of data base. The purpose of the analysis for subchannel blockage in the subassembly of LMR is to represent quantitatively that the maximum damage due to the accident is within the safety criteria. The computational program should be developed to simulate the thermal hydraulic phenomena and to verify the safety of LMR for the accident. For the purpose, the hybrid scheme has been implemented into the MATRA-LMR code based on the upwind scheme to analyze the various flow fields occurred in the subchannel blockage accident. The turbulent mixing models using the CFX code were assessed to compute more precisely the heat transfer between subchannels. Through this assessment, empirical correction factors of 1.7 for the heat conduction, 0.006 for the turbulent mixing coefficient were obtained. The distributed resistance model instead of wire forcing function has been developed to represent the more exact flow field due to wire-wrap. Other models, such as heat conductor model and various turbulent mixing model, have been implemented into the MATRA-LMR. The ORNL THORS 19-Pin FFM-5B tests have been assessed to validate above new models using the improved MATRA-LMR. The results using MATRA-LMR were well agreed with the experimental data. The subchannel blockage accidents which assumed to be occurred at the three locations for the conceptual plant of KALIMER-600 have been analysed according to blockage size using the MATRA-LMR code. The results of calculations for the design basis events which 6 subchannels were blocked showed the margins of the 290 7.dog. C up to the

Recycling option search for a 600 MWE sodium-cooled transmutation fast reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Kyo; Kim, Myung Hyun [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

2015-02-15

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro- SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to 20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

Directory of Open Access Journals (Sweden)

Yong Kyo Lee

2015-02-01

Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

Review of PRA methodology for LMFBR

International Nuclear Information System (INIS)

Yang, J. E.

1999-02-01

Probabilistic Risk Assessment (PRA) has been widely used as a tool to evaluate the safety of NPPs (Nuclear Power Plants), which are in the design stage as well as in operation. Recently, PRA becomes one of the licensing requirements for many existing and new NPPs. KALIMER is a Liquid Metal Fast Breeder Reactor (LMFBR) being developed by KAERI. Since the design concept of KALIMER is similar to that of the PRISM plant developed by GE, it would be appropriate to review the PRA methodology of PRISM as the first step of KALIMER PRA. Hence, in this report summarizes the PRA methodology of PRISM plant, and the required works for the PSA of KALIMER based on the reviewed results. The PRA technology of PRISM plant consists of following five major tasks: (1) development of initiating event list, (2) development of system event tree, (3) development of core response event tree, (4) development of containment response event tree, and (5) consequences and risk estimation. The estimated individual and societal risk measures show that the risk from a PRISM module is substantially less than the NRC goal. Each task is compared to the PRA methodology of Light Water Reactor (LWR)/Pressurized Heavy Water Reactor (PHWR). In the report, each task of PRISM PRA methodology is reviewed and compared to the corresponding part of LWR/PHWR PSA performed in Korea. The parts that are not modeled appropriately in PRISM PRA are identified, and the recommendations for KALIMER PRA are stated. (author). 14 refs., 9 tabs., 4 figs

Dynamic analysis of Korean nuclear fuel cycle with fast reactor systems

International Nuclear Information System (INIS)

Jeong, Chang Joon

2004-12-01

The Korean nuclear fuel cycle scenario was analyzed by the dynamic analysis method, including Pressurized Water Reactor (PWR), Canadian Deuterium Uranium (CANDU) and fast reactor systems. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 1%. After setting up the once-through fuel cycle model, the Korea Advanced Liquid Metal Reactor (KALIMER) scenario was modeled to investigate the fuel cycle parameters. For the analysis of the fast reactor fuel cycle, both KAILMER-150 and KALIMER-600 reactors were considered. In this analysis, the spent fuel inventory as well as the amount of plutonium, Minor Actinides (MA) and Fission Products (FP) of the recycling fuel cycle was estimated and compared to that of the once-through fuel cycle. Results of the once-through fuel cycle calculation showed that the demand grows up to 64 GWe and total amount of spent fuel would be ∼102 kt in 2100. If the KALIMER scenario is implemented, the total spent fuel inventory can be reduced by ∼80%. However it was found that the KALIMER scenario does not contribute to reduce the amount of MA and FP, which is important when designing a repository. For the further destruction of MA, an actinide burner can be considered in the future nuclear fuel cycle

Preliminary Comparative Evaluation Study on Reference Design of GEN-IV SFR

Energy Technology Data Exchange (ETDEWEB)

Sim, Yoon Sub; Kim, Yeong Il; Hong, Ser Gi (and others)

2005-11-15

A fast reactor has a good transmutation capability and it enables breeding of fuel and use of a closed fuel cycle. By these characteristics of a fast reactor, the limited uranium resources of the world can be much more effectively utilized and the nuclear wastes of a high level of radioactivity and toxicity from the current nuclear power reactors of LWRs and HWRs can be drastically reduced in its volume and the management of the wastes can be easily treated. Also electricity can be generated more effectively since a fast reactor has the feature of high operation temperature. These features of a fast reactor makes it inevitable on a long term basis to construct fast reactors in Korea. The domestic fast reactor technology level, however, is at the level of coming out of a beginning stage and needs utilization of international expertise. Recently an international cooperation program called GIF has been formulated and our KALIMER was selected as one of the two reference designs for the international joint R and D works with JSFR of Japan. In the current frame of the GIF program, the two selected reference designs are supposed to be evaluated against each other in future and one design is to be finally selected. To make the international cooperation program directed more useful to our fast reactor technology development, it is required to strengthen the competitiveness of KALIMER so that it can be selected. To meet the necessity, a study was made in this research for pre-evaluation of the GIF reference designs and setting up plans for development of designs and technology that will enhance the competitiveness of KALIMER.

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

Assessment of Proliferation Resistance of Closed Nuclear Fuel Cycle System with Sodium Cooled Fast Reactors Using INPRO Evaluation Methodology

International Nuclear Information System (INIS)

Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

2007-11-01

Using the INPRO methodology, the proliferation resistance of an innovative nuclear energy system(INS) defined as a closed nuclear fuel cycle system consisting of KALIMER and pyroprocessing, has been assessed. Considering a very early development stage of the INS concept, the PR assessment is carried out based on intrinsic features, if required information and data are not available. The PR assessment of KALIMER and JSFR using the INPRO methodology affirmed that an adequate proliferation resistance has been achieved in both INSs CNFC-SFR, considering the assessor's progress and maturity of design development. KALIMER and JSFR are developed or being developed conforming to the targets and criteria defined for developing Gen IV nuclear reactor system. Based on these assessment results, proliferation resistance and physical protection(PR and PP) of KALIMER and JSFR are evaluated from the viewpoint of requirements for future nuclear fuel cycle system. The envisioned INSs CNFC-SFR rely on active plutonium management based on a closed fuel cycle, in which a fissile material is recycled in an integrated fuel cycle facility within proper safeguards. There is no isolated plutonium in the closed fuel cycle. The material remains continuously in a sequence of highly radioactive matrices within inaccessible facilities. The proliferation resistance assessment should be an ongoing analysis that keeps up with the progress and maturity of the design of Gen IV SFR

Assessment of Proliferation Resistance of Closed Nuclear Fuel Cycle System with Sodium Cooled Fast Reactors Using INPRO Evaluation Methodology

Energy Technology Data Exchange (ETDEWEB)

Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

2007-11-15

Using the INPRO methodology, the proliferation resistance of an innovative nuclear energy system(INS) defined as a closed nuclear fuel cycle system consisting of KALIMER and pyroprocessing, has been assessed. Considering a very early development stage of the INS concept, the PR assessment is carried out based on intrinsic features, if required information and data are not available. The PR assessment of KALIMER and JSFR using the INPRO methodology affirmed that an adequate proliferation resistance has been achieved in both INSs CNFC-SFR, considering the assessor's progress and maturity of design development. KALIMER and JSFR are developed or being developed conforming to the targets and criteria defined for developing Gen IV nuclear reactor system. Based on these assessment results, proliferation resistance and physical protection(PR and PP) of KALIMER and JSFR are evaluated from the viewpoint of requirements for future nuclear fuel cycle system. The envisioned INSs CNFC-SFR rely on active plutonium management based on a closed fuel cycle, in which a fissile material is recycled in an integrated fuel cycle facility within proper safeguards. There is no isolated plutonium in the closed fuel cycle. The material remains continuously in a sequence of highly radioactive matrices within inaccessible facilities. The proliferation resistance assessment should be an ongoing analysis that keeps up with the progress and maturity of the design of Gen IV SFR.

Steps to Advanced CANDU 600

International Nuclear Information System (INIS)

Oh, Yongshick; Brooks, G. L.

1988-01-01

The CANDU nuclear power system was developed from merging of AECL heavy water reactor technology with Ontario Hydro electrical power station expertise. The original four units of Ontario Hydro's Pickering Generating Station are the first full-scale commercial application of the CANDU system. AECL and Ontario Hydro then moved to the next evolutionary step, a more advanced larger scale design for four units at the Bruce Generating Station. CANDU 600 followed as a single unit nuclear electric power station design derived from an amalgam of features of the multiple unit Pickering and Bruce designs. The design of the CANDU 600 nuclear steam supply system is based on the Pickering design with improvements derived from the Bruce design. For example, most CANDU 600 auxiliary systems are based on Bruce systems, whereas the fuel handling system is based on the Pickering system. Four CANDU 600 units are in operation, and five are under construction in Romania. For the additional four units at Pickering Generating Station 'B', Ontario Hydro selected a replica of the Pickering 'A' design with limited design changes to maintain a high level of standardization across all eight units. Ontario Hydro applied a similar policy for the additional four units at Bruce Generating Station 'B'. For the four unit Darlington station, Ontario Hydro selected a design based on Bruce with improvements derived from operating experience, the CANDU 600 design and development programs

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

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.

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

International Nuclear Information System (INIS)

Bruschi, H.J.

1996-01-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse's advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

Energy Technology Data Exchange (ETDEWEB)

Bruschi, H.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1996-10-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse`s advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

Geochemical investigation of Sasa tailings dam material and its influence on the Lake Kalimanci surficial sediments (Republic of Macedonia – preliminary study

Directory of Open Access Journals (Sweden)

Petra Vrhovnik

2011-12-01

Full Text Available This research is aimed at investigating the mineralogical characteristics of the tailings material and heavy metal contents of the tailings material deposited close to the Sasa Pb-Zn Mine in the Osogovo Mountains (eastern Macedonia and on its possible impact on Lake Kalimanci. The mineral composition of Sasa Mine tailings materialis dominated by quartz, pyrite, galena, sphalerite, magnetite and others. Geochemical analysis was performed in a certified commercial laboratory for the following elements: Mo, Cu, Pb, Zn, Ni, As, Cd, Sb, Bi, Ag, Al, Fe, Mn, S.Analysis revealed very high concentrations of toxic metals in the tailing material – with average values [ mg kg-1]:Mo 2.9, Cu 279, Pb 3975, Zn 5320, Ni 30, As 69, Cd 84, Sb 4.2, Bi 9.4 and Ag 4.1. The multi-element contamination of Sasa Mine tailings material was assigned a pollution index greater of 15, indicating that the tailings material from Sasa Mine contains very high amounts of toxic metals and represents a high environmental risk for surrounding ecosystems. For this reason the influence of discharged tailings dam material into Lake Kalimanci which liesapproximately 12 km lower than Sasa Mine, was also established. Calculated pollution index values for Lake Kalimancisediments vary from 21 to 65 and for Sasa mine surficial tailings dam material from 15 to 60.

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.

Workshop UNK-600 (proceedings)

International Nuclear Information System (INIS)

Zajtsev, A.M.; Bitykov, S.I.

1994-01-01

Proceedings are presented of the workshop devoted to the accelerating storage complex of IHEP (UNK-600). In the first section is given the information on the present status of the UNK-600 and particle channels design and on the adopted experiment NEPTUN-A. In the papers of the second section are discussed hadron physics investigations at 600 GeV. Experiments in the neutrino and muon beams are analyzed. A possible program of studying the charged kaon rare decays is described

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.

Status of national programmes on fast reactors and accelerator driven systems in Korea

International Nuclear Information System (INIS)

Hahn, Dohee; Kim, Yeong Il

2001-01-01

The LMR (liquid metal cooled reactors) Design Technology Development Project was approved as a national long-term R and D program in 1992 by the Korea Atomic Energy Commission (KAEC). KAEC decided to develop and construct an LMR with the goal of developing an LMR that can serve as a long term power supplier with competitive economics and enhanced safety. Based upon the KAEC decision, the Korea Atomic Energy Research Institute (KAERI) has been developing KALIMER (Korea Advanced Liquid Metal Reactor). According to the revised National Nuclear Energy Promotion Plan of June 1997, the basic design of KALIMER is to be completed by 2006 and feasibility of the construction is to be examined sometime during the mid 2010s. Phase 1 of three years of the LMR Design Technology Development Project was completed in March 2000 and a preliminary conceptual design report has been issued. The conceptual design of KALIMER will be finalized during Phase 2 of the project, which was started in April 2000 and will take two years. KAERI is also carrying out research and development on an accelerator driven system, called HYPER, for the transmutation of nuclear waste and energy production through the transmutation process. The HYPER program is being performed within the framework of the national mid- and long-term nuclear research plan. KAERI aims to develop a system concept and type of roadmap by the year 2001, and to complete conceptual design of the HYPER system by the year 2007. (author)

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.

Design and synthesis of N-(4-aminopyridin-2-yl)amides as B-Raf(V600E) inhibitors.

Science.gov (United States)

Li, Xiaokai; Shen, Jiayi; Tan, Li; Zhang, Zhang; Gao, Donglin; Luo, Jinfeng; Cheng, Huimin; Zhou, Xiaoping; Ma, Jie; Ding, Ke; Lu, Xiaoyun

2016-06-15

B-Raf(V600E) was an effective target for the treatment of human cancers. Based on a pan-Raf inhibitor TAK-632, a series of N-(4-aminopyridin-2-yl)amide derivatives were designed as novel B-Raf(V600E) inhibitors. Detailed structure-activity studies of the compounds revealed that most of the compounds displayed potent enzymatic activity against B-Raf(V600E), and good selectivity over B-Raf(WT). One of the most promising compound 4l exhibited potent inhibitory activity with an IC50 value of 38nM for B-raf(V600E), and displayed antiproliferative activities against colo205 and HT29 cells with IC50 values of 0.136 and 0.094μM, respectively. It also displayed good selectivity on both enzymatic and cellular assays over B-Raf(WT). These inhibitors may serve as lead compounds for further developing novel B-Raf(V600E) inhibitors as anticancer drugs. Copyright © 2016 Elsevier Ltd. All rights reserved.

The State-of-the-Art Report on the Liquid Metal Reactor

International Nuclear Information System (INIS)

Hahn, Do Hee; Kim, Yeong Il; Kim, Seong O; Lee, Jae Han; Lee, Yong Bum

2006-03-01

The status of the sodium cooled metal fuel core technology and the design methodology were described. The preliminary design concepts of the metal fuel core were established for KALIMER. A systematic study on the development fluid and I and C system has been carried out, and the conceptual design of NSSS of the 150MWe and 600MWe LMRs such as the design of PHTS, IHTS, RHRS, SGS and related technology with BOP is established together with the computational technology. The development of detection system of the sodium-water reaction, core blockage and the conceptual design of the control system of large capacity LMR are being carried. The important technological areas for mechanical structure design of LMR are high temperature thin structure design, seismic isolation design, In-Service- Inspection technology, and the economic design. The highlighted performances for the safety analysis were the developments of the containment analysis code CONTAIN-LMR-K, the safety analysis code SSC-K and the flow blockage analysis code. The safety criteria were set up, the safety analysis for the equilibrium core, the HCDA analysis, and the containment performance analysis were performed. The recent SSC-K 1.3 version turns out to be reliable after the indirect verification throughout qualitative/quantitative assessments

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...

Reactivity feedback models for SSC-K

Energy Technology Data Exchange (ETDEWEB)

Han, Do Hee; Kwon, Young Min; Kim, Kyung Du; Chang, Won Pyo [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-06-01

Safety of KALIMER is assured by the inherent safety of the core and passive safety of the safety-related systems. For the safety analysis of a new reactor design such as KALIMER, analysis models, which are consistent with the design, have to be developed for a plant-wide transient and safety analysis code. Efforts for the development of reactivity feedback models for SSC-K, which is now being developed for the safety analysis of KALIMER, is described in this report. Models for Doppler, sodium density/void, fuel axial expansion, core radial expansion, and CRDL expansion have been developed. Test runs have been performed for the unprotected accident for the verification of the models. Use of KALIMER reactivity coefficients and future development of models for GEM and PSDRS would make it possible to analyze the response of KALIMER under TOP as well as LOF and LOHS accident conditions using SSC-K. (author). 5 refs., 64 figs., 2 tabs.

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.

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....

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...

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

Intermediate-break LOCA analyses for the AP600 design

International Nuclear Information System (INIS)

Boyack, B.E.; Lime, J.F.

1995-01-01

A postulated double-ended guillotine break of a direct-vessel-injection line in an AP600 plant has been analyzed. This event is characterized as an intermediate break loss-of-coolant accident (IBLOCA). Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations performed with the TRAC-PFl/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. The key processes occurring in an AP600 during a IBLOCA are primary coolant system depressurization, inventory depletion, inventory replacement via emergency core coolant injection, continuous core cooling, and long-term decay heat rejection to the atmosphere. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated Thus, the observation that the core is continuously cooled should be verified for the latter phase of the long-term cooling period, the interval when sump injection and containment cooling processes are important

Updated TRAC analysis of an 80% double-ended cold-leg break for the AP600 design

International Nuclear Information System (INIS)

Lime, J.F.; Boyack, B.E.

1995-01-01

An updated TRAC 80% large-break loss-of-coolant accident (LBLOCA) has been calculated for the Westinghouse AP600 advanced reactor design, The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The 80% break size was calculated by Westinghouse to be the most severe large-break size for the AP600 design. The LBLOCA transient was calculated to 144 s. Peak cladding temperatures (PCTS) were well below the Appendix K limit of 1,478 K (2,200 F), but very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCT for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown phase. The reasons for these differences are still being investigated. Additional break sizes and break locations need to be analyzed to confirm the most severe break postulated by Westinghouse

INPRO phase 1B (2nd part) joint study

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Bum; Kim, Young In; Hahn, Do Hee (and others)

2006-08-15

In this project, the Korean innovative nuclear energy system(INS) concept was assessed to be contributory to IAEA's INPRO Joint Study on CNFC-FR. The Korean INS concept was defined as an integrated system consisting of a sodium-cooled, metal fueled fast reactor KALIMER and a PWR(including CANDU)-KALIMER coupled closed nuclear fuel cycle for the Joint Study. From the results of the national scenario study performed based o the Korean INS concept, it has veen confirmed that the deployment of KALIMER from 2030 until 2100 could reduce the amount of domestic spent fuel from PWRs and CANDUs with no further increase in PWR spent fuel thereafter. And the amount of minor actinides disposed as high level would be decreased to zero with complete replacement of PWRs with KALIMERs. Based on the results of the national scenario study, a preliminary assessment of the Korean INS concept has been performed using the INPRO methodology and user's manuals. During the INS assessment, items requiring either improvement or complement have been detected in order to dedicate to INPRO's effort to improve the methodology. The INPRO methodology generally lack a consistency in a level of depth and quantity of evaluation criteria and parameters for six areas within the INPRO framework. It needs to complement application methods and guidances applicable to various technology levels as well as illustrations of assessment tools. In addition, it needs to develop quantification and aggregation of evaluated results, application of weighting factor methods, and a synthetic manual for integrated assessment procedure and methodology.

INPRO phase 1B (2nd part) joint study

International Nuclear Information System (INIS)

Lee, Yong Bum; Kim, Young In; Hahn, Do Hee

2006-08-01

In this project, the Korean innovative nuclear energy system(INS) concept was assessed to be contributory to IAEA's INPRO Joint Study on CNFC-FR. The Korean INS concept was defined as an integrated system consisting of a sodium-cooled, metal fueled fast reactor KALIMER and a PWR(including CANDU)-KALIMER coupled closed nuclear fuel cycle for the Joint Study. From the results of the national scenario study performed based o the Korean INS concept, it has veen confirmed that the deployment of KALIMER from 2030 until 2100 could reduce the amount of domestic spent fuel from PWRs and CANDUs with no further increase in PWR spent fuel thereafter. And the amount of minor actinides disposed as high level would be decreased to zero with complete replacement of PWRs with KALIMERs. Based on the results of the national scenario study, a preliminary assessment of the Korean INS concept has been performed using the INPRO methodology and user's manuals. During the INS assessment, items requiring either improvement or complement have been detected in order to dedicate to INPRO's effort to improve the methodology. The INPRO methodology generally lack a consistency in a level of depth and quantity of evaluation criteria and parameters for six areas within the INPRO framework. It needs to complement application methods and guidances applicable to various technology levels as well as illustrations of assessment tools. In addition, it needs to develop quantification and aggregation of evaluated results, application of weighting factor methods, and a synthetic manual for integrated assessment procedure and methodology

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.

Analysis of Accident Scenarios for the Development of Probabilistic Safety Assessment Model for the Metallic Fuel Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Kim, Tae Woon; Park, S. Y.; Yang, J. E.; Kwon, Y. M.; Jeong, H. Y.; Suk, S. D.; Lee, Y. B.

2009-03-01

The safety analysis reports which were reported during the development of sodium cooled fast reactors in the foreign countries are reviewed for the establishment of Probabilistic Safety Analysis models for the domestic SFR which are under development. There are lots of differences in the safety characteristics between the mixed oxide (MOX) fuel SFR and metallic fuel SFR. Metallic fuel SFR is under development in Korea while MOX fuel SFR is under development in France, Japan, India and China. Therefore the status on the development of fast reactors in the foreign countries are reviewed at first and then the safety characteristics between the MOX fuel SFR and the metallic fuel SFR are reviewed. The core damage can be defined as coolant voiding, fuel melting, cladding damage. The melting points of metallic fuel and the MOX fuel is about 1000 .deg. C and 2300 .deg. C, respectively. The high energy stored in the MOX fuel have higher potential to voiding of coolant compared to the possibility in the metallic fuel. The metallic fuel has also inherent reactivity feedback characteristic that the metallic fuel SFR can be shutdown safely in the events of transient overpower, loss of flow, and loss of heat sink without scram. The metallic fuel has, however, lower melting point due to the eutectic formation between the uranium in metallic fuel and the ferrite in metallic cladding. It is needed to identify the core damage accident scenarios to develop Level-1 PSA model. SSC-K computer code is used to identify the conditions in which the core damage can occur in the KALIMER-600 SFR. The accident cases which are analyzed are the triple failure accidents such as unprotected transient over power events, loss of flow events, and loss of heat sink events with impaired safety systems or functions. Through the analysis of the triple failure accidents for the KALIMER-600 SFR, it is found that the PSA model developed for the PRISM reactor design can be applied to KALIMER-600. However

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

Concept and designs of new-generation fast reactors

International Nuclear Information System (INIS)

Mitenkov, F.M.

1993-01-01

This article discusses the general safety requirements and characteristics for future nuclear power plants. It examines various designs - loop, block, and integrated layouts for reactors. Specifically, the article focuses an integrated design for sodium-cooled fast reactors noting that the BN-600 reactor has operated accident-free over the past 12 years. An obvious advantage of this scheme is that the coolant of the primary loop is localized in one volume (in a vessel), there are no short connections and large-diameter pipes, which of course sharply reduces the probability in coolant leaks. With an integrated scheme the problem of embrittlement of the reactor vessel by neutron irradiation is obviated. The neutron fluence for the vessels of the AST-500 and VPBER-600 reactors, built with an integrated scheme, is less than 10 17 cm -2 . Such a fluence does not cause any appreciable change in the mechanical properties of the vessel steel. The integrated layout of the reactor makes it possible to build a containment vessel. In this case it is possible to eliminate the danger of the reactor core drying out and thus cooling of the reactor in emergency situations can be simplified substantially. In an integrated layout, however, access is more difficult to the equipment inside the reactor, thus limiting or complicating maintenance work. The integrated layout, therefore, requires the use of highly reliable equipment built according to designs that have been proven in operation and have been passed representative service-life tests under laboratory conditions. The integrated layout considerably increases the mass and size characteristics of the reactor. New solutions thus are needed for the organization of work on reactor fabrication and assembly. In the case of the BN-600 and Superphenix reactors the welding of the reactor vessels and the assembly work were done on the building site

Preliminary analyses of AP600 using RELAP5

International Nuclear Information System (INIS)

Modro, S.M.; Beelman, R.J.; Fisher, J.E.

1991-01-01

This paper presents results of preliminary analyses of the proposed Westinghouse Electric Corporation AP600 design. AP600 is a two loop, 600 MW (e) pressurized water reactor (PWR) arranged in a two hot leg, four cold leg nuclear steam supply system (NSSS) configuration. In contrast to the present generation of PWRs it is equipped with passive emergency core coolant (ECC) systems. Also, the containment and the safety systems of the AP600 interact with the reactor coolant system and each other in a more integral fashion than present day PWRs. The containment in this design is the ultimate heat sink for removal of decay heat to the environment. Idaho National Engineering Laboratory (INEL) has studied applicability of the RELAP5 code to AP600 safety analysis and has developed a model of the AP600 for the Nuclear Regulatory Commission. The model incorporates integral modeling of the containment, NSSS and passive safety systems. Best available preliminary design data were used. Nodalization sensitivity studies were conducted to gain experience in modeling of systems and conditions which are beyond the applicability of previously established RELAP5 modeling guidelines or experience. Exploratory analyses were then undertaken to investigate AP600 system response during postulated accident conditions. Four small break LOCA calculations and two large break LOCA calculations were conducted

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...

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.

Role of Fugen HWR in Japan and design of a 600 MWe demonstration reactor

International Nuclear Information System (INIS)

Sawai, Sadamu.

1982-03-01

Fugen, a 165 MWe prototype of a heavy water-moderated, boiling light water-cooled reactor, has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work for a 600 MWe demonstration plant has been carried out since 1973. The important systems and components, such as pressure tube assemblies and control rod drive mechanism, are essentially the same as those of Fugen. However, some modification is made owing to the experience obtained in Fugen and LWrs. In the HWR Fugen, plutonium and uranium are effectively used, and plutonium makes the coolant void reactivity more negative, which results in the increase of the stability and safety of the reactor. On August 4, 1981, the ad hoc committee submitted the final report to the Japanese Atomic Energy Commission, in which the construction of a 600 MWe demonstration plant was recommended. As for the research and development on reactor safety, coolant leak detectors, the performance of ECCS, and safety design codes are enumerated. Since 1965, mixed oxide fuel has been developed, and 168 fuel assemblies were loaded in Fugen, but failure did not occur. (Kako, I.)

Preliminary Design and Computational Fluid Dynamics Analysis of Supercritical Carbon Dioxide Turbine Blade

International Nuclear Information System (INIS)

Jeong, Wi S.; Kim, Tae W.; Suh, Kune Y.

2007-01-01

The supercritical gas turbine Brayton cycle has been adopted in the secondary loop of the Generation IV Nuclear Energy Systems, and planned to be installed in power conversion cycles of the nuclear fusion reactors as well. The supercritical carbon dioxide (SCO 2 ) is one of widely considered fluids for this concept. The potential beneficiaries include the Secure Transportable Autonomous Reactor- Liquid Metal (STAR-LM), the Korea Advanced Liquid Metal Reactor (KALIMER) and Battery Omnibus Reactor Integral System (BORIS) which is being developed at the Seoul National University. The reason for these welcomed applications is that the SCO 2 Brayton cycle can achieve higher overall energy conversion efficiency than the steam turbine Rankine cycle. Seoul National University has recently been working on the SCO 2 based Modular Optimized Brayton Integral System (MOBIS). The MOBIS design power conversion efficiency is about 45%. Gas turbine design is crucial part in achieving this high efficiency. In this paper, the preliminary analysis on first stage of gas turbine was performed using CFX as a solver

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.)

Conceptual study of advanced PWR core design

International Nuclear Information System (INIS)

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

1997-09-01

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

Conceptual study of advanced PWR core design

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

Selection of JAERI'S HTGR-GT concept

International Nuclear Information System (INIS)

Muto, Y.; Ishiyama, S.; Shiozawa, S.

2001-01-01

In JAERI, a feasibility study of HTGR-GT has been conducted as an assigned work from STA in Japan since January 1996. So far, the conceptual or preliminary designs of 600, 400 and 300 MW(t) power plants have been completed. The block type core and pebble-bed core have been selected in 600 MW(t) and 400/300 MW(t), respectively. The gas-turbine system adopts a horizontal single shaft rotor and then the power conversion vessel is separated into a turbine vessel and a heat exchanger vessel. In this paper, the issues related to the selection of these concepts are technically discussed. (author)

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

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...

Establishment of design concept of large capacity passive reactor KP1000 and performance evaluation of safety system for LBLOCA

Energy Technology Data Exchange (ETDEWEB)

Kim, Seong O.; Hwang, Young Dong; Kim, Young In; Chang, Moon Hee

1997-03-01

This study was performed to establish the design concepts and to evaluate the performance of safety features of large capacity passive reactor (1000 MWe grade). The design concepts of the large capacity passive reactor `KP1000` were established to generate 1000 MW electric power based on the AP600 of Westinghouse by increasing the number of reactor coolant loop and by increasing the size of reactor internals/core. To implement the analysis of the LBLOCA for KP1000, various kinds of computer codes being considered, it was concluded that RELAP5 was the most appropriate one in availability and operations in present situation. By the analysis of the computer code `RELAP5/Mod3.2.1.2`, following conclusions were derived as described below. First, by spectrum analysis of the discharge factor of the berak part, the most conservative discharge factor C{sub D}=1.2 and the PCT value of KP1000 was 1254F, which is slightly higher than the value of AP600 but is much less than the existing active reactor `Kori 3 and 4` where blowdown PCT value is 1693.4 deg F and reflooding PCT is 1918.4 deg F. Second, after the 200 seconds from the initiation of LBLOCA, IRWST water was supplied in a stable state and the maximum temperature of clad were maintained in a saturated condition. Therefore, it was concluded that the passive safety features of KP1000 keep reactor core from being damaged for large break LOCA. (author). 11 refs., 28 tabs., 37 figs.

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

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

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

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

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

Containment integrity analysis for the (W) advanced AP600

International Nuclear Information System (INIS)

Gagnon, A.F.; Howe, K.S.

1989-01-01

This paper reports that since 1987, Westinghouse has been performing containment cooling analyses in support of the Advanced AP600 plant design. This program was intended to verify the feasibility of the passive containment cooling system features of the AP600 design. To support this design, containment analyses of the AP600 containment for a large break LOCA and a large Steam Line Break were performed. The transient results indicate the feasibility of the passive containment design by demonstrating the capability to remove sufficient heat to limit containment atmosphere conditions to within acceptable limits following these postulated accidents. These results also indicate that the PCCS can reduce containment pressure to less than one-quarter design pressure at 24 hours following the most severe accident scenario thereby minimizing containment leakage concerns

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...

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.

Applicability of RELAP5 for safety analysis of AP600 and PIUS reactors

International Nuclear Information System (INIS)

Motloch, C.G.; Modro, S.M.

1990-01-01

An assessment of the applicability of using RELAP5 for performing safety analyses of the AP600 and PIUS advanced reactor concepts is being performed. This ongoing work is part of a larger safety assessment of advanced reactors sponsored by the United States Nuclear Regulatory Commission. RELAP5 models and correlations are being reviewed from the perspective of the new AP600 and PIUS phenomena and features that could be important to reactor safety. The purpose is to identify those areas in which new mathematical models of physical phenomena would be required to be added to RELAP5. In most cases, the AP600 and PIUS designs and systems and the planned and off-normal operations are similar enough to current Pressurized Water Reactors (PWR) that RELAP5 safety analysis applicability is unchanged. However, for AP600 the single most important systemic and phenomenological difference between it and current PWRs is in the close coupling between the reactor system and the containment during postulated Loss of Coolant Accident (LOCA) events. This close coupling may require the addition of some thermal-hydraulic models to RELAP5. And for PIUS, the most important new feature is the thermal density locks. These and other important safety-related features are discussed. This document presents general descriptions of RELAP5, AP600, and PIUS, describes the new features and phenomena of the reactors, and discusses the code/reactors safety-related issues. 32 refs., 4 figs., 2 tabs

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)

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

Review of the proposed materials of construction for the SBWR and AP600 advanced reactors

International Nuclear Information System (INIS)

Diercks, D.R.; Shack, W.J.; Chung, H.M.; Kassner, T.F.

1994-06-01

Two advanced light water reactor (LWR) concepts, namely the General Electric Simplified Boiling Water Reactor (SBWR) and the Westinghouse Advanced Passive 600 MWe Reactor (AP600), were reviewed in detail by Argonne National Laboratory. The objectives of these reviews were to (a) evaluate proposed advanced-reactor designs and the materials of construction for the safety systems, (b) identify all aging and environmentally related degradation mechanisms for the materials of construction, and (c) evaluate from the safety viewpoint the suitability of the proposed materials for the design application. Safety-related systems selected for review for these two LWRs included (a) reactor pressure vessel, (b) control rod drive system and reactor internals, (c) coolant pressure boundary, (d) engineered safety systems, (e) steam generators (AP600 only), (f) turbines, and (g) fuel storage and handling system. In addition, the use of cobalt-based alloys in these plants was reviewed. The selected materials for both reactors were generally sound, and no major selection errors were found. It was apparent that considerable thought had been given to the materials selection process, making use of lessons learned from previous LWR experience. The review resulted in the suggestion of alternate an possibly better materials choices in a number of cases, and several potential problem areas have been cited

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.

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)

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…

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.

AP600 level of automation: United States utility perspective

International Nuclear Information System (INIS)

Bekkerman, A.Y.

1997-01-01

Design of the AP600 advanced nuclear plant man-machine interface system (M-MIS) is guided by the applicable requirements from the Utility Requirements Document (URD). However, the URD has left certain aspects of the M-MIS to be determined by the designer working together with utilities sponsoring the work. This is particularly true in the case of the level of automation to be designed into the M-MIS. Based on experience from currently operating plants, utilities have specified the identity and roles of personnel in the control room, which has led to establishing a number of level of automation issues for the AP600. The key role of automated computerized procedures in the AP600 automation has been determined and resolved. 5 refs

Preliminary Assessment of PHTS Pump Piping Break Accident of DSFR-600

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Bae, Moohoon; Choi, Yongwon; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

KINS is evaluating the applicability of TRACE code for safety analysis of SFR Since 2012. Based on the steady-state input deck for Demonstration Sodium Cooled Fast Reactor 600MW (DSFR-600) component-wise specific modeling is developed for DSFR-600. Preliminary analysis was performed with TRACE code for DSFR-600 PHTS pump piping break accident. The calculation result showed that the calculated safety parameters are conforms to the design criteria for DBA accidents. RHRS design of DSFR-600 and its performance during transient was also reviewed by sensitivity study on the effect of sodium condition to the transient decay heat removal capability of RHRS. Following insights are identified. These should be considered in improving the design also in licensing review of SFR safety analysis. The transient performance of RHRS might differ from the component's design capacity. RHRS's transient performance also should be included in the design documents and validated with reasonable test and/or analysis with consideration of the variation of coolant conditions during transient. The analytic model used for safety analysis should consider 3-D effect of vessel pool and its uncertainty with reasonable conservatism.

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.

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...

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.

Development of the Biological Experimental Design Concept Inventory (BEDCI)

Science.gov (United States)

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

2014-01-01

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

Structural concepts and details for seismic design

International Nuclear Information System (INIS)

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

1991-01-01

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

Combined raman spectrometer/laser-induced breakdown spectrometer design concept

Science.gov (United States)

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

2017-11-01

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

Applying and incorporating user driven innovation when designing concepts

DEFF Research Database (Denmark)

Thorp Hansen, Claus; Brønnum, Louise

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

Extending Sociotechnical Design to Project Conception

DEFF Research Database (Denmark)

Kampf, Constance Elizabeth

2011-01-01

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

600 MW nuclear power database

International Nuclear Information System (INIS)

Cao Ruiding; Chen Guorong; Chen Xianfeng; Zhang Yishu

1996-01-01

600 MW Nuclear power database, based on ORACLE 6.0, consists of three parts, i.e. nuclear power plant database, nuclear power position database and nuclear power equipment database. In the database, there are a great deal of technique data and picture of nuclear power, provided by engineering designing units and individual. The database can give help to the designers of nuclear power

Concept Generation for Design Creativity A Systematized Theory and Methodology

CERN Document Server

Taura, Toshiharu

2013-01-01

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

Level design concept, theory, and practice

CERN Document Server

Kremers, Rudolf

2009-01-01

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

The workspace design concept: A new framework of participatory ergonomics

OpenAIRE

Broberg, Ole

2007-01-01

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

DESIGNING STUDY FOR A FAMILY FARM WITH 600 GOATS

Directory of Open Access Journals (Sweden)

I. PĂDEANU

2007-10-01

Full Text Available In the hilly and plains area of Banat region, goat rearing for milk production haschances to become a profitable business. After Romania integration into the EUmarket there will be no quotas for goat milk and meat production. Also, importantlow-production arable land areas (over 3 million hectares will be laying fallow inthe next years, spectacularly increasing the fodder area for ruminants. There a fewgoat family farms having an efficient technological flow and with possibilities toprocess the milk in Romania. In this paper the bases are laid down for projecting afarm with 600 indigenous goats, to be exploited in an intensive system andgenetically improved with Sannen or French Alpine he-goats. The followingreproduction indices were planned for the 600 goats: goats in estrus per season96%, fecundity 95%, goats that keep the pregnancy 98%, kidding goats 90%,prolificacy 170%, and birth rate 152 kids for 100 dam goats. The total populationafter weaning the kids is 600 goats, 24 he-goats, and 173 reproduction female kids.For feeding this population 66.8 ha are required out of which 43.1 ha with grassespasture, 2.1 ha alfalfa, 10.2 ha corn, 4.2 ha barley, and 6.6 ha oats. Goats arehoused in 4 shelters, in 12 group pens of 48 heads. Goats will be fed year-round withgrass haylage, oats straw and concentrate mixtures. This farm will produce 2250 Hlmilk per year (mechanical milking, 150 reproduction female kids for selling at 8-9months of age, 500 fattening kids, and 120 culled goats sold for meat. The annuallyestimated gross income will be 34000 EUR.

Borehole disposal design concept

International Nuclear Information System (INIS)

RANDRIAMAROLAHY, J.N.

2007-01-01

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

Concept design on RH maintenance of CFETR Tokamak reactor

International Nuclear Information System (INIS)

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

2014-01-01

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

Design Concepts. Teacher Edition. Marketing Education LAPs.

Science.gov (United States)

Hawley, Jana

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

The Role of Design Concepts in the Development of Industrial Services

DEFF Research Database (Denmark)

Pekkala, Janne; Ylirisku, Salu

2017-01-01

B-to-B industrial manufacturing organisations are moving focus from designing products to services. This transition challenges the management of innovating, which is increasingly collaborative and networked. Organisations need to be able to tackle the related uncertainty in order to prepare, secure......-to-B industrial manufacturing. Eight roles for design concepts are identified in the 11-month study, and these are presented as stories concretising how design concepts functioned. Design concepts were utilised in 1) anticipating future, 2) implementing design, 3) training, 4) engaging in dialogue, 5) setting...... goals, 6) establishing vocabulary in organisation, 7) planning and securing resources, and 8) linking projects....

The workspace design concept: A new framework of participatory ergonomics

DEFF Research Database (Denmark)

Broberg, Ole

2007-01-01

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

Human-centered incubator: beyond a design concept

OpenAIRE

Goossens, R H M; Willemsen, H

2013-01-01

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

Design for All in Scandinavia - a strong concept.

Science.gov (United States)

Bendixen, Karin; Benktzon, Maria

2015-01-01

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

Project Design Concept for Monitoring and Control System

International Nuclear Information System (INIS)

MCGREW, D.L.

2000-01-01

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

Development of a metal-clad advanced composite shear web design concept

Science.gov (United States)

Laakso, J. H.

1974-01-01

An advanced composite web concept was developed for potential application to the Space Shuttle Orbiter main engine thrust structure. The program consisted of design synthesis, analysis, detail design, element testing, and large scale component testing. A concept was sought that offered significant weight saving by the use of Boron/Epoxy (B/E) reinforced titanium plate structure. The desired concept was one that was practical and that utilized metal to efficiently improve structural reliability. The resulting development of a unique titanium-clad B/E shear web design concept is described. Three large scale components were fabricated and tested to demonstrate the performance of the concept: a titanium-clad plus or minus 45 deg B/E web laminate stiffened with vertical B/E reinforced aluminum stiffeners.

Design and Evaluation of Nextgen Aircraft Separation Assurance Concepts

Science.gov (United States)

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

2012-01-01

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

Using a systems engineering process to develop engineered barrier system design concepts

International Nuclear Information System (INIS)

Jardine, L.J.; Short, D.W.

1991-05-01

The methodology used to develop conceptual designs of the engineered barrier system and waste packages for a geologic repository is based on an iterative systems engineering process. The process establishes a set of general mission requirements and then conducts detailed requirements analyses using functional analyses, system concept syntheses, and trade studies identifications to develop preliminary system concept descriptions. The feasible concept descriptions are ranked based on selection factors and criteria and a set of preferred concept descriptions is then selected for further development. For each of the selected concept descriptions, a specific set of requirements, including constraints, is written to provide design guidance for the next and more detailed phase of design. The process documents all relevant waste management system requirements so that the basis and source for the specific design requirements are traceable and clearly established. Successive iterations performed during design development help to insure that workable concepts are generated to satisfy the requirements. 4 refs., 2 figs

Analysis and Multipoint Design of the TCA Concept

Science.gov (United States)

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

1999-01-01

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

Test bed control center design concept for Tank Waste Retrieval Manipulator Systems

International Nuclear Information System (INIS)

Sundstrom, E.; Draper, J.V.; Fausz, A.

1995-01-01

This paper describes the design concept for the control center for the Single Shell Tank Waste Retrieval Manipulator System test bed and the design process behind the concept. The design concept supports all phases of the test bed mission, including technology demonstration, comprehensive system testing, and comparative evaluation for further development and refinement of the TWRMS for field operations

Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

International Nuclear Information System (INIS)

Suk, H.C; Lee, J.C.; Suh, K.S.; Yuk, K.E.; Whang, W.; Park, J.S.; Eim, J.S.; Bang, K.H.; Eim, M.S.; Rim, C.S.

1982-01-01

The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

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

International Nuclear Information System (INIS)

Szabo, I.

1995-01-01

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

Gallium Electromagnetic (GEM) Thrustor Concept and Design

Science.gov (United States)

Polzin, Kurt A.; Markusic, Thomas E.

2006-01-01

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

Objectives and status of development of AC600

International Nuclear Information System (INIS)

Zhao Chengkun

1997-01-01

AC600 is a medium power capability nuclear power station of next generation, which is developed based on world nuclear power improving tendency, requirements of custom with considering China situation and technical foundation. Its main technical characteristics are as following: advanced core and passive safety system, double loop standard design and international popular equipment. Meanwhile, it a simplification of present system, using advanced control room and pattern construction thus developed the operation reliability of nuclear power station, lower construction and operating cost. In order to accelerate the development of next generation advanced reactor, cooperating with Westinghouse Electric Corporation, the joint economic technical research has been established. Based on AC600, the CAP600 is developed on further improving safety and reliability, economical and electric network adoption of AC600

Conceptual design of a commercial accelerator driven thorium reactor

International Nuclear Information System (INIS)

Fuller, C. G.; Ashworth, R. W.

2010-01-01

This paper describes the substantial work done in underpinning and developing the concept design for a commercial 600 MWe, accelerator driven, thorium fuelled, lead cooled, power producing, fast reactor. The Accelerator Driven Thorium Reactor (ADTR TM) has been derived from original work by Carlo Rubbia. Over the period 2007 to 2009 Aker Solutions commissioned this concept design work and, in close collaboration with Rubbia, developed the physics, engineering and business model. Much has been published about the Energy Amplifier concept and accelerator driven systems. This paper concentrates on the unique physics developed during the concept study of the ADTR TM power station and the progress made in engineering and design of the system. Particular attention is paid to where the concept design has moved significantly beyond published material. Description of challenges presented for the engineering and safety of a commercial system and how they will be addressed is included. This covers the defining system parameters, accelerator sizing, core and fuel design issues and, perhaps most importantly, reactivity control. The paper concludes that the work undertaken supports the technical viability of the ADTR TM power station. Several unique features of the reactor mean that it can be deployed in countries with aspirations to gain benefit from nuclear power and, at 600 MWe, it fits a size gap for less mature grid systems. It can provide a useful complement to Generation III, III+ and IV systems through its ability to consume actinides whilst at the same time providing useful power. (authors)

Design Process for Integrated Concepts with Responsive Building Elements

DEFF Research Database (Denmark)

Aa, Van der A.; Heiselberg, Per

2008-01-01

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

Technical feasibility and reliability of passive safety systems of AC600

International Nuclear Information System (INIS)

Niu, W.; Zeng, X.

1996-01-01

The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished by the Nuclear Power Institute of China. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also

Technical feasibility and reliability of passive safety systems of AC600

Energy Technology Data Exchange (ETDEWEB)

Niu, W; Zeng, X [Nuclear Power Inst. of China, Chendu (China)

1996-12-01

The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also involved. (author). 3 figs, 1 tab.

Role of Fugen-HWR in Japan and design of a 600 MWe demonstration reactor

International Nuclear Information System (INIS)

Sawai, S.

1982-01-01

Fugen, a 165 MWe prototype of a heavy water moderated boiling light water cooled reactor; has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work of the 600 MWe demonstration plant has been carried out since 1973. Important system and components, such as pressure tube assemblies, control rod drive mechanism, etc., are essentially the same as those of Fugen. Some modifications, however, are made especially from the stand point of experiences In the Fugen-HWR, plutonium and uranium would be effectively used; and plutonium could make the coolant void reactivity more negative which would give good results in increasing the reactor stability and safety. On the other hand, nuclear power plants are mainly consisted of LWRs in Japan. Considering the above situations, the Fugen-HWR, coupled with LWRs, is now considered in Japan to contribute to our energy security by using plutonium and depleted uranium extracted from spent fuels of LWRs: thereby reducing the demands On August 4, 1981, the ad hoc committee on the 600 MWe demonstration Fugen-HWR submitted the final report to the Japan AEC, after having had discussions and evaluations. In the report, the ad hoc committee recommended to build the 600 MWE demonstration plant with appropriate supports of the Government. The Japan AEC will be expected to make her decision on the program in the near future. As for the reactor safety R and C, development has been stressed on coolant leak detectors and ECCS performances or Since 1965, many development works have been done for mixed oxide fuel assemblies, both for establishment of the fabrication technology and for clarification of irradiation performances. 196 mixed oxide fuel assemblies have been manufactured for Fugen. 168 of them were loaded and 92 were withdrawn. No fuel has been failured yet. (author)

AP600 containment purge radiological analysis

Energy Technology Data Exchange (ETDEWEB)

O`Connor, M.; Schulz, J.; Tan, C. [Bechtel Power Corporation (United States)] [and others

1995-02-01

The AP600 Project is a passive pressurized water reactor power plant which is part of the Design Certification and First-of-a-Kind Engineering effort under the Advanced Light Water Reactor program. Included in this process is the design of the containment air filtration system which will be the subject of this paper. We will compare the practice used by previous plants with the AP600 approach to meet the goals of industry standards in sizing the containment air filtration system. The radiological aspects of design are of primary significance and will be the focus of this paper. The AP600 Project optimized the design to combine the functions of the high volumetric flow rate, low volumetric flow rate, and containment cleanup and other filtration systems into one multi-functional system. This achieves a more simplified, standardized, and lower cost design. Studies were performed to determine the possible concentrations of radioactive material in the containment atmosphere and the effectiveness of the purge system to keep concentrations within 10CFR20 limits and within offsite dose objectives. The concentrations were determined for various reactor coolant system leakage rates and containment purge modes of operation. The resultant concentrations were used to determine the containment accessibility during various stages of normal plant operation including refueling. The results of the parametric studies indicate that a dual train purge system with a capacity of 4,000 cfm per train is more than adequate to control the airborne radioactivity levels inside containment during normal plant operation and refueling, and satisfies the goals of ANSI/ANS-56.6-1986 and limits the amount of radioactive material released to the environment per ANSI/ANS 59.2-1985 to provide a safe environment for plant personnel and offsite residents.

Design Concepts and Design Practices in Policy-Making and Public Management

DEFF Research Database (Denmark)

Junginger, Sabine

2012-01-01

National governments around the globe are actively seeking new ways to engage in social innovation and are investing in innovation labs and innovation centers where methods and principles of design are now being explored and applied to problems of transforming and innovating the public sector (cf...... governments but they also pose new challenges for policy-makers and public administrators who are not yet familiar with design concepts, principles and methods beyond problem-solving. Despite the many linkages between and among design, designing, policy-making and policy implementation, we have yet to clarify...

Heavy ion driven LMF design concept

International Nuclear Information System (INIS)

Lee, E.P.

1991-08-01

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

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 flood estimation in ungauged basins: probabilistic extension of the design-storm concept

Science.gov (United States)

Berk, Mario; Špačková, Olga; Straub, Daniel

2016-04-01

Design flood estimation in ungauged basins is an important hydrological task, which is in engineering practice typically solved with the design storm concept. However, neglecting the uncertainty in the hydrological response of the catchment through the assumption of average-recurrence-interval (ARI) neutrality between rainfall and runoff can lead to flawed design flood estimates. Additionally, selecting a single critical rainfall duration neglects the contribution of other rainfall durations on the probability of extreme flood events. In this study, the design flood problem is approached with concepts from structural reliability that enable a consistent treatment of multiple uncertainties in estimating the design flood. The uncertainty of key model parameters are represented probabilistically and the First-Order Reliability Method (FORM) is used to compute the flood exceedance probability. As an important by-product, the FORM analysis provides the most likely parameter combination to lead to a flood with a certain exceedance probability; i.e. it enables one to find representative scenarios for e.g., a 100 year or a 1000 year flood. Possible different rainfall durations are incorporated by formulating the event of a given design flood as a series system. The method is directly applicable in practice, since for the description of the rainfall depth-duration characteristics, the same inputs as for the classical design storm methods are needed, which are commonly provided by meteorological services. The proposed methodology is applied to a case study of Trauchgauer Ach catchment in Bavaria, SCS Curve Number (CN) and Unit hydrograph models are used for modeling the hydrological process. The results indicate, in accordance with past experience, that the traditional design storm concept underestimates design floods.

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

Science.gov (United States)

Lanz, Amelia; Davis, Rebecca G

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

Advanced Spacesuit Portable Life Support System Packaging Concept Mock-Up Design & Development

Science.gov (United States)

O''Connell, Mary K.; Slade, Howard G.; Stinson, Richard G.

1998-01-01

A concentrated development effort was begun at NASA Johnson Space Center to create an advanced Portable Life Support System (PLSS) packaging concept. Ease of maintenance, technological flexibility, low weight, and minimal volume are targeted in the design of future micro-gravity and planetary PLSS configurations. Three main design concepts emerged from conceptual design techniques and were carried forth into detailed design, then full scale mock-up creation. "Foam", "Motherboard", and "LEGOtm" packaging design concepts are described in detail. Results of the evaluation process targeted maintenance, robustness, mass properties, and flexibility as key aspects to a new PLSS packaging configuration. The various design tools used to evolve concepts into high fidelity mock ups revealed that no single tool was all encompassing, several combinations were complimentary, the devil is in the details, and, despite efforts, many lessons were learned only after working with hardware.

Review of the Technical Status on the Debris Bed Cooling Model

Energy Technology Data Exchange (ETDEWEB)

Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

2007-09-15

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris0.

Review of the Technical Status on the Debris Bed Cooling Model

International Nuclear Information System (INIS)

Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

2007-09-01

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris

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

Science.gov (United States)

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

2015-01-01

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

Soft shell hard core concept for aircraft impact resistant design

International Nuclear Information System (INIS)

Chen, C.; Rieck, P.J.

1978-01-01

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

An Artificial-Gravity Space-Settlement Ground-Analogue Design Concept

Science.gov (United States)

Dorais, Gregory A.

2016-01-01

The design concept of a modular and extensible hypergravity facility is presented. Several benefits of this facility are described including that the facility is suitable as a full-scale artificial-gravity space-settlement ground analogue for humans, animals, and plants for indefinite durations. The design is applicable as an analogue for on-orbit settlements as well as those on moons, asteroids, and Mars. The design creates an extremely long-arm centrifuge using a multi-car hypergravity vehicle travelling on one or more concentric circular tracks. This design supports the simultaneous generation of multiple-gravity levels to explore the feasibility and value of and requirements for such space-settlement designs. The design synergizes a variety of existing technologies including centrifuges, tilting trains, roller coasters, and optionally magnetic levitation. The design can be incrementally implemented such that the facility can be operational for a small fraction of the cost and time required for a full implementation. Brief concept of operation examples are also presented.

Design and simulation experimental study of bracket plates in steam generator for AC600 PWR

International Nuclear Information System (INIS)

Zhang Fuyuan; Zhang Wenqi; Ji Quankai; Zeng Xi; Xie Yongyao

1998-01-01

Seven-holes type bracket plate at the inlet nozzle and three-holes taper bracket plate at outlet nozzle are designed. According to 'local form and structure change' simulation theory, hydraulic models and simulators for the simulative experiments are designed. Taking water as the medium, the simulative experiments have been completed at the room temperature. The ζ-Re curves (here, ζ is the local pressure loss coefficient at the nozzles after the bracket plates are installed and Re is Reynolds number) have been got. Based on the experimental results, the computation and the analysis have been shown that. If the bracket plates are used in the steam generator (SG) of AC600 PWR, the pressure drop of primary side in the SG is about 14 percent higher than that of the 55/19 B style SG

Long-term plutonium storage: Design concepts

International Nuclear Information System (INIS)

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

1994-01-01

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

Design Concepts for Cooled Ceramic Composite Turbine Vane

Science.gov (United States)

Boyle, Robert J.; Parikh, Ankur H.; Nagpal, VInod K.

2015-01-01

The objective of this work was to develop design concepts for a cooled ceramic vane to be used in the first stage of the High Pressure Turbine(HPT). To insure that the design concepts were relevant to the gas turbine industry needs, Honeywell International Inc. was subcontracted to provide technical guidance for this work. The work performed under this contract can be divided into three broad categories. The first was an analysis of the cycle benefits arising from the higher temperature capability of Ceramic Matrix Composite(CMC) compared with conventional metallic vane materials. The second category was a series of structural analyses for variations in the internal configuration of first stage vane for the High Pressure Turbine(HPT) of a CF6 class commercial airline engine. The third category was analysis for a radial cooled turbine vanes for use in turboshaft engine applications. The size, shape and internal configuration of the turboshaft engine vanes were selected to investigate a cooling concept appropriate to small CMC vanes.

Design concept of HYPER (HYbrid Power Extraction Reactor)

International Nuclear Information System (INIS)

Park, Won S.; Song, Tae Y.; Yu, Dong H.; Kim, Chang H.

1999-01-01

Korea Atomic Energy Research Institute (KAERI) has been performing accelerator driven system related research and development called HYPER for the transmutation of nuclear waste and energy production through the transmutation process. Some major design features of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with a minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Pb-Bi is adopted as a coolant and spallation target material. 1 GeV 13 mA proton beam is designed to be provided for HYPER. HYPER is to transmute about 380 kg of TRU a year and produce 1000 MWth power. The support ratio of HYPER for LWR units producing the same power is believed to be 5 to 6. (author)

Application of Sensitivity Analysis in Design of Integrated Building Concepts

DEFF Research Database (Denmark)

Heiselberg, Per; Brohus, Henrik; Hesselholt, Allan Tind

2007-01-01

analysis makes it possible to identify the most important parameters in relation to building performance and to focus design and optimization of integrated building concepts on these fewer, but most important parameters. The sensitivity analyses will typically be performed at a reasonably early stage...... the design requirements and objectives. In the design of integrated building concepts it is beneficial to identify the most important design parameters in order to more efficiently develop alternative design solutions or more efficiently perform an optimization of the building performance. The sensitivity...

Current fusion power plant design concepts

International Nuclear Information System (INIS)

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

1976-09-01

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

Extending Sociotechnical design to project conception

DEFF Research Database (Denmark)

Kampf, Constance

2009-01-01

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

Design of a multivariable controller for a CANDU 600 MWe nuclear power plant using the INA method

International Nuclear Information System (INIS)

Roy, N.; Boisvert, J.; Mensah, S.

1984-04-01

The development of large and complex nuclear and process plants requires high-performance control systems, designed with rigorous multivariable techniques. This work is part of an analytical study demonstrating the real potential of multivariable methods. It covers every step in the design of a multi-variable controller for a Gentilly-2 type CANDU 600 MWe nuclear power plant using the Inverse Nyquist Array (INA) method. First the linear design model and its preliminary modifications are described. The design tools are reviewed and the operations required to achieve open-loop diagonal dominance are thoroughly described. Analysis of the closed-loop system is then performed and a feedback matrix is selected to meet the design specifications. The performance of the controller on the linear model is verified by simulation. Finally, the controller is implemented on the reference non-linear model to assess its overall performance. The results show that the INA method can be used successfully to design controllers for large and complex systems

Development of preliminary design concept for a multifunction display and control system for the Orbiter crew station. Task 4: Design concept recommendation

Science.gov (United States)

Spiger, R. J.; Farrell, R. J.; Holcomb, G. A.

1982-01-01

Application of multifunction display and control systems to the NASA Orbiter spacecraft offers the potential for reducing crew workload and improving the presentation of system status and operational data to the crew. A design concept is presented for the application of a multifunction display and control system (MFDCS) to the Orbital Maneuvering System and Electrical Power Distribution and Control System on the Orbiter spacecraft. The MFDCS would provide the capability for automation of procedures, fault prioritization and software reconfiguration of the MFDCS data base. The MFDCS would operate as a stand-alone processor to minimize the impact on the current Orbiter software. Supervisory crew command of all current functions would be retained through the use of several operating modes in the system. Both the design concept and the processes followed in defining the concept are described.

Research on green building design based on ecological concept

Directory of Open Access Journals (Sweden)

Zhang Ping Qing

2016-01-01

Full Text Available At present, the protection of the ecological environment and the promotion of green building has been recognized and widely promoted.With the rapid development of the construction industry, Architecture design will inevitably require the resentation of its unique form and charm to reflect the ecological concept and ecological culture, because of the unique nature of the art and the particularity of the environment. To establish the ecological concept of green building design and vigorously develop the green green building has a complementary role to alleviate the pressure on resources,and to speed up the eco city planning design, and to realize the sustainable development of the city, and to protect the urban ecological environmental.

High Altitude Venus Operations Concept Trajectory Design, Modeling and Simulation

Science.gov (United States)

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

2015-01-01

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

Project Design Concept - Primary Ventilation System

International Nuclear Information System (INIS)

MCGREW, D.L.

2000-01-01

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

Benefits of Low Boron Core Design Concept for PWR

Energy Technology Data Exchange (ETDEWEB)

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

2009-10-15

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

Benefits of Low Boron Core Design Concept for PWR

International Nuclear Information System (INIS)

Daing, Aung Tharn; Kim, Myung Hyun

2009-01-01

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

Engineering reliability in design phase: An application to AP-600 reactor passive safety system

International Nuclear Information System (INIS)

Majumdr, D.; Siahpush, A.S.; Hills, S.W.

1992-01-01

A computerized reliability enhancement methodology is described that can be used at the engineering design phase to help the designer achieve a desired reliability of the system. It can take into account the limitation imposed by a constraint such as budget, space, or weight. If the desired reliability of the system is known, it can determine the minimum reliabilities of the components, or how many redundant components are needed to achieve the desired reliability. This methodology is applied to examine the Automatic Depressurization System (ADS) of the new passively safe AP-600 reactor. The safety goal of a nuclear reactor dictates a certain reliability level of its components. It is found that a series parallel valve configuration instead of the parallel-series configuration of the four valves in one stage would improve the reliability of the ADS. Other valve characteristics and arrangements are explored to examine different reliability options for the system

Development of environmentally advanced hydropower turbine system design concepts

Energy Technology Data Exchange (ETDEWEB)

Franke, G.F.; Webb, D.R.; Fisher, R.K. Jr. [Voith Hydro, Inc. (United States)] [and others

1997-08-01

A team worked together on the development of environmentally advanced hydro turbine design concepts to reduce hydropower`s impact on the environment, and to improve the understanding of the technical and environmental issues involved, in particular, with fish survival as a result of their passage through hydro power sites. This approach brought together a turbine design and manufacturing company, biologists, a utility, a consulting engineering firm and a university research facility, in order to benefit from the synergy of diverse disciplines. Through a combination of advanced technology and engineering analyses, innovative design concepts adaptable to both new and existing hydro facilities were developed and are presented. The project was divided into 4 tasks. Task 1 investigated a broad range of environmental issues and how the issues differed throughout the country. Task 2 addressed fish physiology and turbine physics. Task 3 investigated individual design elements needed for the refinement of the three concept families defined in Task 1. Advanced numerical tools for flow simulation in turbines are used to quantify characteristics of flow and pressure fields within turbine water passageways. The issues associated with dissolved oxygen enhancement using turbine aeration are presented. The state of the art and recent advancements of this technology are reviewed. Key elements for applying turbine aeration to improve aquatic habitat are discussed and a review of the procedures for testing of aerating turbines is presented. In Task 4, the results of the Tasks were assembled into three families of design concepts to address the most significant issues defined in Task 1. The results of the work conclude that significant improvements in fish passage survival are achievable.

Development of environmentally advanced hydropower turbine system design concepts

International Nuclear Information System (INIS)

Franke, G.F.; Webb, D.R.; Fisher, R.K. Jr.

1997-08-01

A team worked together on the development of environmentally advanced hydro turbine design concepts to reduce hydropower''s impact on the environment, and to improve the understanding of the technical and environmental issues involved, in particular, with fish survival as a result of their passage through hydro power sites. This approach brought together a turbine design and manufacturing company, biologists, a utility, a consulting engineering firm and a university research facility, in order to benefit from the synergy of diverse disciplines. Through a combination of advanced technology and engineering analyses, innovative design concepts adaptable to both new and existing hydro facilities were developed and are presented. The project was divided into 4 tasks. Task 1 investigated a broad range of environmental issues and how the issues differed throughout the country. Task 2 addressed fish physiology and turbine physics. Task 3 investigated individual design elements needed for the refinement of the three concept families defined in Task 1. Advanced numerical tools for flow simulation in turbines are used to quantify characteristics of flow and pressure fields within turbine water passageways. The issues associated with dissolved oxygen enhancement using turbine aeration are presented. The state of the art and recent advancements of this technology are reviewed. Key elements for applying turbine aeration to improve aquatic habitat are discussed and a review of the procedures for testing of aerating turbines is presented. In Task 4, the results of the Tasks were assembled into three families of design concepts to address the most significant issues defined in Task 1. The results of the work conclude that significant improvements in fish passage survival are achievable

Computational analysis of supercritical carbon dioxide flow around a turbine and compressor BLADE

International Nuclear Information System (INIS)

Kim, Tae W.; Kim, Nam H.; Suh, Kune Y.; Kim, Seung O.

2007-01-01

The turbine and compressor isentropic efficiencies are one of the major parameters affecting the overall Brayton cycle efficiency. Thus, the optimal turbine and compressor design should contribute to the economics of future nuclear fission and fusion energy systems. A computation analysis was performed utilizing CFX for the supercritical carbon dioxide (SCO 2 ) flow around a turbine and compressor blade to check on the potential efficiency of the turbine and compressor which determine such basic design values as the blade (or impeller) and nozzle (or diffuser) types, blade height, and minimum and maximum radii of the hub and tip. Basic design values of the turbine and compressor blades based on the Argonne National Laboratory (ANL) design code was generated by ANSYS BladeGen TM . The boundary conditions were based on the KALIMER-600 secondary loop. Optimal SCO 2 turbine and compressor blades were developed for high efficiency of 90% by the computational analysis. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-11-12

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Mechatronical Aided Concept (MAC) in Intelligent Transport Vehicles Design

OpenAIRE

Pavel Pavlasek

2003-01-01

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

The Ductile Design Concept for Seismic Actions in Miscellaneous Design Codes

Directory of Open Access Journals (Sweden)

M. Budescu

2009-01-01

Full Text Available The concept of ductility estimates the capacity of the structural system and its components to deform prior to collapse, without a substantial loss of strength, but with an important energy amount dissipated. Consistent with the „Applied Technology Council” (ATC-34, from 1995, it was agreed that the reduction seismic response factor to decrease the design force. The purpose of this factor is to transpose the nonlinear behaviour of the structure and the energy dissipation capacity in a simplified form that can be used in the design stage. Depending on the particular structural model and the design standard the used values are different. The paper presents the characteristics of the ductility concept for the structural system. Along with this the general way of computing the reserve factor with the necessary explanations for the parameters that determine the behaviour factor are described. The purpose of this paper is to make a comparison between different international norms for the values and the distribution of the behaviour factor. The norms from the following countries are taken into consideration: the United States of America, New Zealand, Japan, Romania and the European general seismic code.

Mechatronical Aided Concept (MAC in Intelligent Transport Vehicles Design

Directory of Open Access Journals (Sweden)

Pavel Pavlasek

2003-01-01

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

Toroidal field coil design concept and structural support system for CTHR

Energy Technology Data Exchange (ETDEWEB)

Chianese, R. B.; Kelly, J. L.; Ruck, G. W.

1980-09-01

The CTHR conceptual design consists of a magnetically confined (tokamak) fusion reactor fitted with a fertile uranium blanket. The fusion driver concept was based on an ignited plasma. All concepts and parameters were selected on the basis that technical feasibility would be achieved by 1995 to assure a viable commercial operation in the early to mid-21st century. The reactor was designed to achieve good fissile fuel production, with electricity production being a second order priority. However, the resulting concepts that evolved were all excellent power producers which significantly improved the economic performance. The subsystems discussed in the following paragraphs provide a background of the application for the TF coil design described in this report.

Toroidal field coil design concept and structural support system for CTHR

International Nuclear Information System (INIS)

Chianese, R.B.; Kelly, J.L.; Ruck, G.W.

1980-09-01

The CTHR conceptual design consists of a magnetically confined (tokamak) fusion reactor fitted with a fertile uranium blanket. The fusion driver concept was based on an ignited plasma. All concepts and parameters were selected on the basis that technical feasibility would be achieved by 1995 to assure a viable commercial operation in the early to mid-21st century. The reactor was designed to achieve good fissile fuel production, with electricity production being a second order priority. However, the resulting concepts that evolved were all excellent power producers which significantly improved the economic performance. The subsystems discussed in the following paragraphs provide a background of the application for the TF coil design described in this report

Design concept for α-hydrogen-substituted nitroxides.

Science.gov (United States)

Amar, Michal; Bar, Sukanta; Iron, Mark A; Toledo, Hila; Tumanskii, Boris; Shimon, Linda J W; Botoshansky, Mark; Fridman, Natalia; Szpilman, Alex M

2015-02-06

Stable nitroxides (nitroxyl radicals) have many essential and unique applications in chemistry, biology and medicine. However, the factors influencing their stability are still under investigation, and this hinders the design and development of new nitroxides. Nitroxides with tertiary alkyl groups are generally stable but obviously highly encumbered. In contrast, α-hydrogen-substituted nitroxides are generally inherently unstable and rapidly decompose. Herein, a novel, concept for the design of stable cyclic α-hydrogen nitroxides is described, and a proof-of-concept in the form of the facile synthesis and characterization of two diverse series of stable α-hydrogen nitroxides is presented. The stability of these unique α-hydrogen nitroxides is attributed to a combination of steric and stereoelectronic effects by which disproportionation is kinetically precluded. These stabilizing effects are achieved by the use of a nitroxide co-planar substituent in the γ-position of the backbone of the nitroxide. This premise is supported by a computational study, which provides insight into the disproportionation pathways of α-hydrogen nitroxides.

Design, performance, and calculated error of a Faraday cup for absolute beam current measurements of 600-MeV protons

International Nuclear Information System (INIS)

Beck, S.M.

1975-04-01

A mobile self-contained Faraday cup system for beam current measurments of nominal 600-MeV protons was designed, constructed, and used at the NASA Space Radiation Effects Laboratory. The cup is of reentrant design with a length of 106.7 cm and an outside diameter of 20.32 cm. The inner diameter is 15.24 cm and the base thickness is 30.48 cm. The primary absorber is commercially available lead hermetically sealed in a 0.32-cm-thick copper jacket. Several possible systematic errors in using the cup are evaluated. The largest source of error arises from high-energy electrons which are ejected from the entrance window and enter the cup. A total systematic error of -0.83 percent is calculated to be the decrease from the true current value. From data obtained in calibrating helium-filled ion chambers with the Faraday cup, the mean energy required to produce one ion pair in helium is found to be 30.76 +- 0.95 eV for nominal 600-MeV protons. This value agrees well, within experimental error, with reported values of 29.9 eV and 30.2 eV

Design, performance, and calculated error of a Faraday cup for absolute beam current measurements of 600-MeV protons

International Nuclear Information System (INIS)

Beck, S.M.

1975-04-01

A mobile self-contained Faraday cup system for beam current measurements of nominal 600 MeV protons was designed, constructed, and used at the NASA Space Radiation Effects Laboratory. The cup is of reentrant design with a length of 106.7 cm and an outside diameter of 20.32 cm. The inner diameter is 15.24 cm and the base thickness is 30.48 cm. The primary absorber is commercially available lead hermetically sealed in a 0.32-cm-thick copper jacket. Several possible systematic errors in using the cup are evaluated. The largest source of error arises from high-energy electrons which are ejected from the entrance window and enter the cup. A total systematic error of -0.83 percent is calculated to be the decrease from the true current value. From data obtained in calibrating helium-filled ion chambers with the Faraday cup, the mean energy required to produce one ion pair in helium is found to be 30.76 +- 0.95 eV for nominal 600 MeV protons. This value agrees well, within experimental error, with reported values of 29.9 eV and 30.2 eV. (auth)

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)

Preliminary ALARA design concept for SMART

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

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

Preliminary ALARA design concept for SMART

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

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

Preliminary ALARA design concept for SMART

International Nuclear Information System (INIS)

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

1999-03-01

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

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)

Design concepts and status of the Korean next generation reactor (KNGR)

International Nuclear Information System (INIS)

Cho, Sung Jae; Kim, Han Gon

1999-01-01

The national project to develop KNGR, a 4000 MWth evolutionary advanced light water reactor (ALWR), has been organized in three phases according to the development status in 1992. During the first phase, the top-tier design requirements and the design concepts to meet the requirements had been established. The project is currently in the second phase of which the major objective is to complete the basic design sufficient to confirm the plant safety. This paper describes the overall design concepts and status of the KNGR briefly which developed and/or being developed through the project. (author)

Current Status of the Transmutation Reactor Technology and Preliminary Evaluation of Transmutation Performance of the KALIMER Core

Energy Technology Data Exchange (ETDEWEB)

Hong, Ser Gi; Sim, Yoon Sub; Kim, Yeong Il; Kim, Young Gyum; Lee, Byung Woon; Song, Hoon; Lee, Ki Bog; Jang, Jin Wook; Lee, Dong Uk

2005-08-15

devised. It has been considered that the degradations of core performances resulting from increase of the transmutation rate are very important problems. From the analysis results of the state-of-art of the nuclear transmutation technology, the following technical research topics are determined as the technical solution ways for the future development and enhancement of the transmutation technology; 1) the improvement of core safety through the reduction of the coolant void reactivity worth by using the void duct assembly, 2) the design of a reference transmutation reactor for the future transmutation research through the change of the KALIMER-600 reactor core into the transmutation reactor and its core performance analysis, 3) the optimization study of the hybrid loading of uranium-free fuel and uranium fuel to improve the transmutation rate and the core safety parameters. Finally, the feasibility of the transmutation core suggested above where the void duct assemblies are devised to improve the sodium void reactivity worth and to achieve the power flattening under a single fuel enrichment and a single type of fuel assembly is analyzed and assessed. The results show that this core has its sodium coolant void reactivity less than 3$ and this core can transmutate the TRU nuclides discharged from two LWRs of the same thermal power.

Second generation waste package design and storage concept for the Yucca Mountain Repository

International Nuclear Information System (INIS)

Armijo, Joseph Sam; Kar, Piyush; Misra, Manoranjan

2006-01-01

The reference waste package design and operating mode to be used in the Yucca Mountain Repository is reviewed. An alternate (second generation) operating concept and waste package design is proposed to reduce the risk of localized corrosion of waste packages and to reduce repository costs. The second generation waste package design and storage concept is proposed for implementation after the initial licensing and operation of the reference repository design. Implementation of the second generation concept at Yucca Mountain would follow regulatory processes analogous to those used successfully to extend the design life and uprate the power of commercial light water nuclear reactors in the United States. The second generation concept utilizes the benefits of hot dry storage to minimize the potential for localized corrosion of the waste package by liquid electrolytes. The second generation concept permits major reductions in repository costs by increasing the number of fuel assemblies stored in each waste package, by eliminating the need for titanium drip shields and by fabricating the outer container from corrosion resistant low alloy carbon steel

Periodic Virtual Cell Manufacturing (P-VCM) - Concept, Design and Operation

NARCIS (Netherlands)

Slomp, Jannes; Krushinsky, Dimitry; Caprihan, Rahul

2011-01-01

This paper presents and discusses the concept of Periodic Virtual Cell Manufacturing (P-VCM). After giving an illustrative example of the operation and design complexity of a P-VCM system, we present an industrial case to study the applicability of the concept. The illustrative example and the

Prophylactic single-dose administration of 600 mg clindamycin versus 4-time administration of 600 mg clindamycin in orthognathic surgery: A prospective randomized study in bilateral mandibular sagittal ramus osteotomies

NARCIS (Netherlands)

Lindeboom, Jerôme A. H.; Baas, Eric M.; Kroon, Frans H. M.

2003-01-01

Objective. The purpose of this study was to compare a single 600-mg dose of preoperative intravenously administered clindamycin with a 24-hour 600-mg regimen of clindamycin as prophylaxis for postoperative infections in bilateral sagittal ramus osteotomies. Study design. Seventy patients were

Systems analysis and futuristic designs of advanced biofuel factory concepts.

Energy Technology Data Exchange (ETDEWEB)

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

2007-10-01

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

600 a Current Leads with Dry and Compact Warm Terminals

CERN Document Server

Andersen, T P; Vullierme, B

2002-01-01

For the LHC magnet test benches 26 pairs of conventional helium vapour-cooled 600 A current leads are required. The first pair of 600 A current leads has been designed and built by industry and tested at CERN. The main component of the lead is the heat exchanger, which consists of two concentric copper pipes. Special attention was also given to the design of the warm terminal in order to avoid any condensation and to resist at an electrical test of 2 kV. The paper describes construction details and compares calculated and measured values of the main parameters.

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)

FE-study for lithostatic pressure measurement in the 600 m borehole experiment

International Nuclear Information System (INIS)

Hamilton, L.F.M.; Benneker, P.B.J.M.

1990-05-01

In the Asse-2 salt mine an experiment is set up by ECN in the 600 m borehole to perform in-situ convergence measurements which can be used to validate or to determine the constitutive relations between stresses and deformations of the rocksalt. An experiment is planned in which the convergence of the borehole can be measured with different pressures created in the borehole. For this experiment a device has been developed at ECN which also will be used to measure the in-situ elasticity of the salt. This measuring device is designed in such a way that a pressure can be realized in the borehole and the deformation of the hole can be measured at the same time. In this report analyses are presented that are used to adjust the design of the pressure unit to the specific needs induced by the fact that the depth of the borehole is only 300 m due to drilling problems instead of the intended 600 m. Since the lithostatic pressure at this depth is smaller the convergence rate of the borehole is reduced. From the results presented in this report it can be concluded that it is not necessary to change the basic concept of the measurement as it was planned in the 600 m deep borehole. After the device has been placed into the borehole at a distance of 3 m from the bottom the empty space must be filled up with salt concrete up to 3 m above the pressure unit. In this way the interaction with the borehole bottom and the transition between the open and the filled borehole can be neglected. Some changes in the design are necessary to be able to measure the deformations of the borehole with enough accuracy. Based on these changes a measuring program has been defined in such a way that the experimental period is optimally used and the expected evolution of the volume change and pressure can be measured with enough accuracy. For three different constitutive models a prediction is made for the evolution of the pressure and the volume change resulting from the defined measuring program

Divertor remote handling for DEMO: Concept design and preliminary FMECA studies

Energy Technology Data Exchange (ETDEWEB)

Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Di Gironimo, G. [ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

2015-10-15

Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor mover: hydraulic telescopic boom concept design. • An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • FMECA studies started on the DEMO divertor mover. - Abstract: The paper describes a concept design of a remote handling (RH) system for replacing divertor cassettes and cooling pipes in future DEMO fusion power plant. In DEMO reactor design important considerations are the reactor availability and reliable maintenance operations. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative designs of the end effector to grip and manipulate the divertor cassette are presented in this work. Both concepts are hydraulically actuated, based on ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. Taking advantage of the ITER RH background and experience, the proposed hydraulic RH system is compared with the rack and pinion system currently designed for ITER and is an object of simulations at Divertor Test Platform (DTP2) in VTT's Labs of Tampere, Finland. Pros and cons will be put in evidence.

The Role of Design Concepts in the Development of Industrial Services

DEFF Research Database (Denmark)

Pekkala, Janne; Ylirisku, Salu

2017-01-01

B-to-B industrial manufacturing organisations are moving focus from designing products to services. This transition challenges the management of innovating, which is increasingly collaborative and networked. Organisations need to be able to tackle the related uncertainty in order to prepare, secure......, and plan their use of resources. Design concepts are known to have various beneficial roles in product and service development in various development contexts. In this article we study how design concepts were utilised within, and between, three development projects in a Finnish company in the context of B...

Comparative design of structures concepts and methodologies

CERN Document Server

Lin, Shaopei

2016-01-01

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

76 FR 18960 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2011-04-06

... B4-600, B4-600R, and F4-600R Series Airplanes, and Model C4-605R Variant F Airplanes (Collectively... July 20, 2005; have been performed in service. (2) Airbus Model A300 B4-605R, B4-622R, F4-605R, and F4... C4-600R, and A300 F4-600R series airplanes (fitted with a trim tank), all serial numbers, except...

Core design with respect to the safety concept

International Nuclear Information System (INIS)

Kollmar, W.

1981-01-01

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

Lithium-thionyl chloride battery design concepts for maximized power applications

Science.gov (United States)

Kane, P.; Marincic, N.

The need for primary batteries configured to deliver maximized power has been asserted by many different procuring activities. Battery Engineering Inc. has developed some specific design concepts and mastered some specialized techniques utilized in the production of this type of power source. The batteries have been successfully bench tested during the course of virtually all of these programs, with ultimate success coming in the form of two successful test launches under the USAF Plasma Effects Decoy Program. This paper briefly discusses some of these design concepts and the rationale behind them.

AP600 passive containment cooling system phenomena identification and ranking table

International Nuclear Information System (INIS)

Spencer, D.R.; Woodcock, Joel

1999-01-01

This paper presents the Phenomena Identification and Ranking Table (PIRT) used in the containment Design Basis Analysis (DBA) for the AP600 nuclear power plant. The PIRT is a tool generally applied to best estimate thermal hydraulic analyses. In the conservative analytical modeling approach used for the AP600 DBA containment pressure response, the PIRT was a tool used to show completeness and relevance of the test database in accordance with the Code of Federal Regulations for advanced plant design. Additionally, the ranking of phenomena by relative importance in a PIRT allows appropriate focusing of resources during model development and licensing review. The focus of the paper is on the organization and structure of the PIRT to show level of detail and format accepted for the AP600, for potential application to other containment designs or accident scenarios. Conclusions of general interest are discussed regarding table organization and structure, the process for developing relative ranking and incorporating expert opinion, and the definition and usage of the relative ranking in support of the conservative evaluation model. The AP600 containment evaluation model approach, as influenced by the relative rankings, is briefly described to put into context this unique application of the PIRT to a conservative methodology. The bases for relative ranking of each phenomenon, which included expert opinion, and quantitative results of scaling and testing, was submitted to the NRC as part of AP600-specific evaluations. Since a PIRT supports the sufficiency of both a testing program and analytical modeling, the process followed to generate and confirm the PIRT, an important part of the licensing acceptance, was a focus of extensive NRC review. General descriptions of key phenomena are provided to aid in understanding the containment PIRT for more general applications for containment evaluations of other PWR designs or for other scenarios. (author)

Retrievable storage concept designs. Final report

International Nuclear Information System (INIS)

Nickell, R.E.

1979-01-01

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

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.

TRAC analysis of an 80% pump-side, cold-leg, large-break loss-of-coolant accident for the Westinghouse AP600 advanced reactor design

International Nuclear Information System (INIS)

Lime, J.F.; Boyack, B.E.

1996-01-01

An updated TRAC 80% pump-side, cold-leg, large-break (LB) loss-of-coolant accident (LOCA) has been calculated for the Westinghouse AP600 advanced reactor design. The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The break size and location were calculated by Westinghouse to be the most severe LBLOCA for the AP600 design. The LBLOCA transient was calculated to 280 s, which is the time of in-containment refueling water-storage-tank injection. All fuel rods were quenched completely by 240 s. Peak cladding temperatures (PCTs) were well below the licensing limit of 1,478 K (2,200 F) but were very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCTs for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown and refill periods. The reasons for these differences are still being investigated

New Design Concept for Universal CCD Controller

Directory of Open Access Journals (Sweden)

Wonyong Han

1994-06-01

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

Design Concepts for Muon-Based Accelerators

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-01

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

Ex-vessel core catcher design requirements and preliminary concepts evaluation

International Nuclear Information System (INIS)

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

1974-01-01

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

75 FR 60611 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2010-10-01

... Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and Model A300 C4...; Model A300 B4-601, B4- 603, B4-620, B4-622, B4-605R, B4-622R, F4-605R, F4-622R, and C4-605R Variant F...-- Dated-- A300 series airplanes......... A300-32A0447..... April 22, 2004. A300 B4-600, B4-600R, and F4...

From corporate control concept to ERP system design

NARCIS (Netherlands)

Kerssens-van Drongelen, I.C.

2003-01-01

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

AIFTDS-8000 - A next generation PCM system: Concept through final design

Science.gov (United States)

Trover, William F.

The development of a new modular PCM system composed of nineteen different types of functional modules is reported. The system is based on the loaf-of-bread packaging concept eliminating the classical fixed size box. The successful design of this packaging concept has been made possible by the building and testing of proof-of-concept models. Thermally driven PC payouts using multilayer PC boards with copper planes for power distribution and heat transfer are essential in achieving the high-end operating temperature of 85 C with a significant margin of safety. The modularity of the design permits low-cost periodic upgrades of key system elements by slice replacement without obsolescence of the majority of the hardware.

Target/blanket conceptual design for the Los Alamos ATW concept

International Nuclear Information System (INIS)

Ames, K.; Cappiello, M.; Ireland, J.; Sapir, J.; Farnum, G.

1992-01-01

The Los Alamos Accelerator Transmutation of Waste (ATW) concept has many potential applications that include defense waste transmutation, defense material production (i.e., tritium and 238 Pu), and the transmutation of hazardous nuclear wastes from commercial nuclear reactors (fission products and actinides). A more advanced long-term Los Alamos effort is investigating the potential of an accelerator- driven system to produce fission energy with a minimal nuclear waste stream. All applications employ a high-energy (800- to 1600-MeV), high-current (25--250 mA) proton linear accelerator as the driver. In this report, we discuss only the target/blanket conceptual design for the commercial nuclear waste application. A conceptual design for the target/blanket of the Los Alamos ATW concept has been presented. The neutronics, mechanical design, and heat transfer have been investigated in some detail for the base-case design. Much more work needs to be done, but at this point it appears that the design is feasible and will approach the design goal of supporting two commercial power reactors with each target/blanket module

The Westinghouse AP600 an advanced nuclear option for small or medium electricity grids

International Nuclear Information System (INIS)

Bruschi, H. J.; Novak, V.

1996-01-01

During the early days of commercial nuclear power, many countries looking to add nuclear power to their energy mix required large plants to meet the energy needs of rapidly growing populations and large industrial complexes. The majority of plants worldwide are in the range of 100 megawatts and beyond. During the 1970s, it became apparent that a smaller nuclear plants would appeal to utilities looking to add additional power capacity to existing grids, or to utilities in smaller countries which were seeking efficient, new nuclear generation capacity for the first time. For instance, the Westinghouse-designed 600 megawatt Krsko plant in Slovenia began operation in 1980, providing electricity to inhabitants of relatively small, yet industrial populations of Slovenia and Croatia. This plant design incorporated the best, proven technology available at that time, based on 20 years of Westinghouse PWR pioneering experience. Beginning in the early 1980s, Westinghouse began to build further upon that experience - in part through the advanced light water reactor programs established by the Electric Power Research institute (EPRI) and the U.S. Department of Energy (DOE) - to design a simplified, advanced nuclear reactor in the 600 megawatt range. Originally, Westinghouse's development of its AP600 (advanced, passive 600-megawatt) plants was geared towards the needs of U.S. utilities which specified smaller, simplified nuclear options for the decades ahead. It soon became evident that the small and medium sized electricity grids of international markets could benefit from this new reactor. From the earliest days of Westinghouse's AP600 development, the corporation invited members of the international nuclear community to take part in the design, development and testing of the AP600 - with the goal of designing a reactor that would meet the diverse needs of an international industry composed of countries with similar, yet different, concerns. (author)

Scaling analysis for the OSU AP600 test facility (APEX)

International Nuclear Information System (INIS)

Reyes, J.N.

1998-01-01

In this paper, the authors summarize the key aspects of a state-of-the-art scaling analysis (Reyes et al. (1995)) performed to establish the facility design and test conditions for the advanced plant experiment (APEX) at Oregon State University (OSU). This scaling analysis represents the first, and most comprehensive, application of the hierarchical two-tiered scaling (H2TS) methodology (Zuber (1991)) in the design of an integral system test facility. The APEX test facility, designed and constructed on the basis of this scaling analysis, is the most accurate geometric representation of a Westinghouse AP600 nuclear steam supply system. The OSU APEX test facility has served to develop an essential component of the integral system database used to assess the AP600 thermal hydraulic safety analysis computer codes. (orig.)

21 CFR 558.600 - Tiamulin.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Tiamulin. 558.600 Section 558.600 Food and Drugs... Animal Feeds § 558.600 Tiamulin. (a) Specifications. Type A article containing 5, 10, or 113.4 grams of tiamulin (as tiamulin hydrogen fumarate) per pound. (b) Approvals. See No. 058198 in § 510.600(c) of this...

Advanced Technology Subsonic Transport Study: N+3 Technologies and Design Concepts

Science.gov (United States)

Raymer, Daniel P.; Wilson, Jack; Perkins, H. Douglas; Rizzi, Arthur; Zhang, Mengmeng; RamirezPuentes, Alfredo

2011-01-01

Conceptual Research Corporation, the Science of the Possible, has completed a two-year study of concepts and technologies for future airliners in the 180-passenger class. This NASA-funded contract was primarily focused on the ambitious goal of a 70 percent reduction in fuel consumption versus the market-dominating Boeing 737-800. The study is related to the N+3 contracts awarded in 2008 by NASA s Aeronautics Research Mission Directorate to teams led by Boeing, GE Aviation, MIT, and Northrop Grumman, but with more modest goals and funding. CRC s contract featured a predominant emphasis on propulsion and fuel consumption, but since fuel consumption depends upon air vehicle design as much as on propulsion technology, the study included notional vehicle design, analysis, and parametric studies. Other NASA goals including NOx and noise reduction are of long-standing interest but were not highlighted in this study, other than their inclusion in the propulsion system provided to CRC by NASA. The B-737-800 was used as a benchmark, parametric tool, and design point of departure. It was modeled in the RDS-Professional aircraft design software then subjected to extensive parametric variations of parasitic drag, drag-due-to-lift, specific fuel consumption, and unsized empty weight. These studies indicated that the goal of a 70 percent reduction in fuel consumption could be attained with roughly a 30 percent improvement in all four parameters. The results were then fit to a Response Surface and coded for ease of use in subsequent trade studies. Potential technologies to obtain such savings were identified and discussed. More than 16 advanced concept designs were then prepared, attempting to investigate almost every possible emerging concept for application to this class airliner. A preliminary assessment of these concepts was done based on their total wetted area after design normalization of trimmed maximum lift. This assessment points towards a Tailless Airliner concept which

Preliminary Development of Regulatory PSA Models for SFR

International Nuclear Information System (INIS)

Choi, Yong Won; Shin, Andong; Bae, Moohoon; Suh, Namduk; Lee, Yong Suk

2013-01-01

Well developed PRA methodology exists for LWR (Light Water Reactor) and PHWR (Pressurized Heavy Water Reactor). Since KAERI is developing a prototype SFR targeting to apply for a license by 2017, KINS needs to have a PRA models to assess the safety of this prototype reactor. The purpose of this study is to develop the regulatory PSA models for the independent verification of the SFR safety. Since the design of the prototype SFR is not mature yet, we have tried to develop the preliminary models based on the design data of KAERI's previous SFR design. In this study, the preliminary initiating events of level 1 internal event for SFR were selected through reviews of existing PRA (LWR, PRISM, ASTRID and KALIMER-600) models. Then, the event tree for each selected initiating event was developed. The regulatory PRA models of SFR developed are preliminary in a sense, because the prototype SFR design is not mature and provided yet. Still it might be utilized for the forthcoming licensing review in assessing the risk of safety issues and the configuration control of the design

Design of Concept of Sustainable Marketing Communication Strategy for a Ideal Industrial Enterprise and Practical Applications of this Concept

Science.gov (United States)

Šujaková, Monika; Golejová, Simona; Sakál, Peter

2017-09-01

In the contribution the authors deal with the design and use of a sustainable marketing communication strategy of an ideal industrial enterprise in the Slovak Republic. The concept of an ideal enterprise is designed to increase the enterprise's sustainable competitiveness through the formation of a corporate image. In the framework of the research, the practical application of the draft concept was realized through a semi-structured interview in the form of propositional logic.

RASC-AL (Revolutionary Aerospace Systems Concepts-Academic Linkage): 2002 Advanced Concept Design Presentation

Science.gov (United States)

2002-01-01

The Revolutionary Aerospace Systems Concepts-Academic Linkage (RASC-AL) is a program of the Lunar and Planetary Institute (LPI) in collaboration with the Universities Space Research Association's (USRA) ICASE institute through the NASA Langley Research Center. The RASC-AL key objectives are to develop relationships between universities and NASA that lead to opportunities for future NASA research and programs, and to develop aerospace systems concepts and technology requirements to enable future NASA missions. The program seeks to look decades into the future to explore new mission capabilities and discover what's possible. NASA seeks concepts and technologies that can make it possible to go anywhere, at anytime, safely, reliably, and affordably to accomplish strategic goals for science, exploration, and commercialization. University teams were invited to submit research topics from the following themes: Human and Robotic Space Exploration, Orbital Aggregation & Space Infrastructure Systems (OASIS), Zero-Emissions Aircraft, and Remote Sensing. RASC-AL is an outgrowth of the HEDS-UP (University Partners) Program sponsored by the LPI. HEDS-UP was a program of the Lunar and Planetary Institute designed to link universities with NASA's Human Exploration and Development of Space (HEDS) enterprise. The first RASC-AL Forum was held November 5-8, 2002, at the Hilton Cocoa Beach Oceanfront Hotel in Cocoa Beach, Florida. Representatives from 10 university teams presented student research design projects at this year's Forum. Each team contributed a written report and these reports are presented.

A mechanism for proven technology foresight for emerging fast reactor designs and concepts

International Nuclear Information System (INIS)

Anuar, Nuraslinda; Muhamad Pauzi, Anas

2016-01-01

The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the β min is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the β min , resulting in a list of candidate designs that possess the β value that is larger than the β min . The proposed methodology can also be applied to purposes other than technological foresight

A mechanism for proven technology foresight for emerging fast reactor designs and concepts

Energy Technology Data Exchange (ETDEWEB)

Anuar, Nuraslinda, E-mail: nuraslinda@uniten.edu.my; Muhamad Pauzi, Anas, E-mail: anas@uniten.edu.my [College of Engineering, Universiti Tenaga Nasional, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia)

2016-01-22

The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the β{sub min} is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the β{sub min}, resulting in a list of candidate designs that possess the β value that is larger than the β{sub min}. The proposed methodology can also be applied to purposes other than technological foresight.

Experimental and calculating substantiation of reactivity balance and energy-release distribution in BN-600 core

International Nuclear Information System (INIS)

Moiseev, A.V.; Khomyakov, Yu.S.; Surov, S.V.

2013-01-01

This paper presents the results of experimental and theoretical work done in 2003-2010 years on substantiation of neutron-physical characteristics of the BN-600 core. 1. Transition to the new core 01M2 with high burnup 11.2% h.a. (the 4-th upgrade of the BN-600 core). Transfer was made without changing the constructive of the core almost by reducing conservatism of design decisions. 2. The end of BN-600 design life cycle and extending it to 10-15 years. Need for analysis and comprehension of the BN-600 experience. 3. Development and introduction of new methods of analysis (precision method of Monte Carlo). 4. In the experiments was a change of equipment and measurement techniques

A design concept of underground facilities for the deep geologic disposal of spent fuel

International Nuclear Information System (INIS)

Lee, Jong Youl; Choi, Heui Joo; Choi, Jong Won; Hahn, Pil Soo

2005-01-01

Spent nuclear fuel from nuclear power plants can be disposed in the underground repository. In this paper, a concept of Korean Reference HLW disposal System (KRS-1) design is presented. Though no site for the underground repository has been specified in Korea, but a generic site with granitic rock is considered for reference spent fuel repository design. To implement the concept, design requirements such as spent fuel characteristics and capacity of the repository and design principles were established. Then, based on these requirements and principles, a concept of the disposal process, the facilities and the layout of the repository was developed

The Triton: Design concepts and methods

Science.gov (United States)

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

1992-01-01

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

New Approach to Concept Feasibility and Design Studies for Astrophysics Missions

Science.gov (United States)

Deutsch, M. J.; McLaughlin, W.; Nichols, J.

1998-01-01

JPL has assembled a team of multidisciplinary experts with corporate knowledge of space mission and instrument development. The advanced Concept Design Team, known as Team X, provides interactive design trades including cost as a design parameter, and advanced visualization for pre-Phase A Studies.

Agent And Component Object Framework For Concept Design Modeling Of Mobile Cyber Physical Systems

Science.gov (United States)

2018-03-01

base design, service-oriented architecture (SOA) and enterprise architecture , brought a new emphasis on business processes and business organization...there are some useful concepts that can be leveraged into an MIGVS architecture . The concept of modeling operational or business behavior logic as...Design 1. Explicit meta model for architecture concepts and relationships 2. Support business or operational modeling and associated events 3

Taking a Concept to Commercialization: Designing Relevant Tests to Address Safety.

Science.gov (United States)

Ferrara, Lisa A

2016-04-01

Taking a product from concept to commercialization requires careful navigation of the regulatory pathway through a series of steps: (A) moving the idea through proof of concept and beyond; (B) evaluating new technologies that may provide added value to the idea; (C) designing appropriate test strategies and protocols; and (D) evaluating and mitigating risks. Moving an idea from the napkin stage of development to the final product requires a team effort. When finished, the product rarely resembles the original design, but careful steps throughout the product life cycle ensure that the product meets the vision.

Development of Core Heat Removal Objective Provision Trees for Sodium-Cooled Fast Reactor Defense-in-Depth Evaluation

Energy Technology Data Exchange (ETDEWEB)

Yang, Huichang; Kang, Bongsuk; Lee, Youngho [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

Based on the definition of Defense-in-Depth levels and safety functions for KALIMER sodium-cooled fast reactor, suggested in the reference and, OPTs for level 1, 2, and 3 defense-in-depth and core heat removal safety function, were developed and suggested in this paper. The purpose of this OPT is first to assure the defensein-depth design during the licensing of Sodium-Cooled Fast Reactors (SFR), but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of PGSFR (Prototype Gen-IV SFR) is not mature yet, the OPT is developed for KALIMER design. Developed OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defensein-depth evaluation frame for the regulatory reviews for the licensing process. At this moment, the identified provisions have both aspects as requirements and design features already adopted in KALIMER design. In the next stage of this study, derived provisions to be adopted will be compared with the actual design features and findings can be suggested as recommendations for the safety improvement.

Design and analysis of advanced flight planning concepts

Science.gov (United States)

Sorensen, John A.

1987-01-01

The objectives of this continuing effort are to develop and evaluate new algorithms and advanced concepts for flight management and flight planning. This includes the minimization of fuel or direct operating costs, the integration of the airborne flight management and ground-based flight planning processes, and the enhancement of future traffic management systems design. Flight management (FMS) concepts are for on-board profile computation and steering of transport aircraft in the vertical plane between a city pair and along a given horizontal path. Flight planning (FPS) concepts are for the pre-flight ground based computation of the three-dimensional reference trajectory that connects the city pair and specifies the horizontal path, fuel load, and weather profiles for initializing the FMS. As part of these objectives, a new computer program called EFPLAN has been developed and utilized to study advanced flight planning concepts. EFPLAN represents an experimental version of an FPS. It has been developed to generate reference flight plans compatible as input to an FMS and to provide various options for flight planning research. This report describes EFPLAN and the associated research conducted in its development.

SCWR Concepts in Japan

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-08-15

Two SCWR concepts are being developed in Japan, one corresponding to the thermal spectrum reactor and the other to the fast spectrum reactor. Yamada et al. described the thermal-spectrum reactor concept referred to as the Japan SCWR (or JSCWR). This concept was developed under the financial support of the Ministry of Economy, Trade and Industry (METI). The basic philosophy of the JSCWR development is to utilize proven light water reactor and supercritical fossil-fired power plant technologies as much as possible to minimize the R&D cost, time and risks. Therefore, the JSCWR is designed as a thermal neutron spectrum reactor using light water as moderator and reactor coolant. The JSCWR plant consists of a pressure-vessel type, once-through reactor and a direct Rankine cycle system. Reactor coolant fed through inlet nozzles is heated up in the core and flows through outlet nozzles with no recirculation in the vessel. Other options to the JSCWR core design are being investigated at the University of Tokyo. The electric output of the JSCWR is assumed to range from 600 MWe to 1700 MWe class to fulfill user’s requirements as much as possible. In this section, the reference value is selected to 1725 MWe, which corresponds to a reactor thermal output of 4039 MWth. Nakatsuka et al. described the core design for the fast-spectrum reactor, which is based on a similar plant system compared to that of the thermal-spectrum reactor. The fast-spectrum reactor, however, would produce higher power rating than the thermal-spectrum one of the same reactor pressure-vessel size. Since the fast-spectrum reactor does not require the moderator, its unit capital cost would be lower than the thermal-spectrum reactor.

Safeguards by design - The early consideration of safeguards concepts

International Nuclear Information System (INIS)

Killeen, T.; Moran, B.; Pujol, E.

2009-01-01

Full-text: The IAEA Department of Safeguards is in the process of formalizing its approach to long-range strategic planning. As a result of this activity new endeavours are being identified. One of these endeavours is to develop a concept known as Safeguards by Design. Safeguarding nuclear material and facilities can be made more effective and cost efficient by improving the safeguardability of the system. By taking into account design features that facilitate the implementation of international safeguards early in the design phase, a concept known as safeguards by design, the proliferation resistance of the system can be improved. This improvement process requires an understanding by designers and operators of safeguards and its underlying principles. To advance the safeguards by design approach, the IAEA determined that there is a need to develop written guidance. This guidance would help the major stakeholders - the designers, operators, owners, and regulatory bodies - to better understand how a facility could be designed, built and operated in such a way that effective safeguards could be implemented at reduced cost and with minimal burden to facility operations. By enlisting the cooperation of Member States through the support programme structure, the IAEA is working to first develop a document that describes the basic principles of safeguards, and the fundamental design features and measures that facilitate the implementation of international safeguards. Facility-specific guidance will then be developed utilizing the resources, expertise and experience of the IAEA and its Member States. This paper will review the foundation for the development of this task, describe the progress that has been made and outline the path forward. (author)

44 CFR 17.600 - Purpose.

Science.gov (United States)

2010-10-01

... 44 Emergency Management and Assistance 1 2010-10-01 2010-10-01 false Purpose. 17.600 Section 17.600 Emergency Management and Assistance FEDERAL EMERGENCY MANAGEMENT AGENCY, DEPARTMENT OF HOMELAND SECURITY GENERAL GOVERNMENTWIDE REQUIREMENTS FOR DRUG-FREE WORKPLACE (GRANTS) § 17.600 Purpose. (a) The...

Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

Energy Technology Data Exchange (ETDEWEB)

Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

1995-09-01

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

76 FR 19724 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2011-04-08

... B4-600, B4-600R, and F4-600R Series Airplanes, and Model C4-605R Variant F Airplanes (Collectively... F4-605R and F4-622R airplanes, and Model A300 C4-605R Variant F airplanes; and Model A310-203, -204...

76 FR 6581 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2011-02-07

... B4-600, B4-600R, and F4-600R Series Airplanes, and Model C4-605R Variant F Airplanes (Collectively...-605R, B4-622R, F4-605R, F4-622R, and C4-605R Variant F airplanes, certificated in any category, all...

Investigating the Act of Design in Discharge Concept Using PMRI

Science.gov (United States)

Lestariningsih; Anwar, Muhammad; Setiawan, Agus Mulyanto

2015-01-01

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

40 CFR 600.313-01 - Timetable for data and information submittal and review.

Science.gov (United States)

2010-07-01

... constraints (except for manufacturers designated under § 600.312(a)(4) who shall submit the information no... running changes (as required under § 600.314(b)), the submission must be made at least five working days before the date of implementation of the running change. (b) A manufacturer may not proceed with any...

40 CFR 600.313-86 - Timetable for data and information submittal and review.

Science.gov (United States)

2010-07-01

... constraints (except for manufacturers designated under § 600.312(a)(4) who shall submit the information no... running changes (as required under § 600.314(b)), the submission must be made at least five working days before the date of implementation of the running change. (b) A manufacturer may not proceed with any...

Design concept for vessels and heat exchangers

International Nuclear Information System (INIS)

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

1981-01-01

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

Evaluation of Standard Concepts Design of Library Interior Physical Environment

Directory of Open Access Journals (Sweden)

Debri Harindya Putri

2018-01-01

Full Text Available Currently the function of a room is not only used as a shelter, the function of the room itself to be increased as a refreshing or relaxation area for users to follow the development of creativity and technology in the field of design. The comfortable factor becomes the main factor that indicates a successful process of creating a space. No exception library. The nature of library seemed stiff because of its function as a place to read, now can be developed and made into more dynamic with the special design concepts or color patterns used. Libraries can be created a special concept that suits the characteristics of the users themselves. Most users of the library, especially in college libraries are teenagers. Naturally, teenagers like to gather with their friends and we have to facilitate this activity in our library design concept. In addition we can also determine the needs of users through research by questionnaire method. The answers of users can be mapped and drawn conclusions. To explore the research, the author reviewed some literature about library interior design and observed the library of Ma Chung University as a case study. The combined results of the method can be concluded and the discovery of ideal standards of physical environment. So, the library can be made as a comfortable reading environment so as to increased interest in reading behavior and the frequent visits of students in the library

Design Concepts of Polycarbonate-Based Intervertebral Lumbar Cages: Finite Element Analysis and Compression Testing

Directory of Open Access Journals (Sweden)

J. Obedt Figueroa-Cavazos

2016-01-01

Full Text Available This work explores the viability of 3D printed intervertebral lumbar cages based on biocompatible polycarbonate (PC-ISO® material. Several design concepts are proposed for the generation of patient-specific intervertebral lumbar cages. The 3D printed material achieved compressive yield strength of 55 MPa under a specific combination of manufacturing parameters. The literature recommends a reference load of 4,000 N for design of intervertebral lumbar cages. Under compression testing conditions, the proposed design concepts withstand between 7,500 and 10,000 N of load before showing yielding. Although some stress concentration regions were found during analysis, the overall viability of the proposed design concepts was validated.

12 CFR 308.600 - Scope.

Science.gov (United States)

2010-01-01

... 12 Banks and Banking 4 2010-01-01 2010-01-01 false Scope. 308.600 Section 308.600 Banks and Banking FEDERAL DEPOSIT INSURANCE CORPORATION PROCEDURE AND RULES OF PRACTICE RULES OF PRACTICE AND PROCEDURE Removal, Suspension, and Debarment of Accountants From Performing Audit Services § 308.600 Scope...

7 CFR 1220.600 - Act.

Science.gov (United States)

2010-01-01

... 7 Agriculture 10 2010-01-01 2010-01-01 false Act. 1220.600 Section 1220.600 Agriculture... CONSUMER INFORMATION Procedures To Request a Referendum Definitions § 1220.600 Act. Act means the Soybean, Promotion, Research, and Consumer Information Act set forth in title XIX, subtitle E, of the Food...

28 CFR 600.5 - Staff.

Science.gov (United States)

2010-07-01

... 28 Judicial Administration 2 2010-07-01 2010-07-01 false Staff. 600.5 Section 600.5 Judicial Administration OFFICES OF INDEPENDENT COUNSEL, DEPARTMENT OF JUSTICE GENERAL POWERS OF SPECIAL COUNSEL § 600.5 Staff. A Special Counsel may request the assignment of appropriate Department employees to assist the...

Design concepts to enhance nuclear power plant protection

International Nuclear Information System (INIS)

Ericson, D.M. Jr.; Varnado, G.B.

1980-01-01

Using a modern design for a nuclear power plant as a point of departure, this study examines the enhancement of protection which may be achieved by changes to the design. These changes include concepts such as complete physical separation of redundant trains of safety equipment, hardened enclosures for water storage tanks, and hardened shutdown heat removal systems. The degree of enhancement (value) is examined in terms such as the potential reduction in the number of vital areas and the increase in probability of adversary sequence interruption. The impacts considered include constraints imposed upon operations and maintenance personnel and increased capital and operating costs. The study concludes that structural design changes alone do not provide significant increases in protection

Concept design of CFETR superconducting magnet system based on different maintenance ports

International Nuclear Information System (INIS)

Zheng, Jinxing; Liu, Xufeng; Song, Yuntao; Wan, Yuanxi; Li, Jiangang; Wu, Sontao; Wan, Baonian; Ye, Minyou; Wei, Jianghua; Xu, Weiwei; Liu, Sumei; Weng, Peide; Lu, Kun; Luo, Zhengping

2013-01-01

Highlights: • This article discussed the concept design of the magnet system of CFETR based on different maintenance port cases. • The major and minor radius of plasma is 5.7 m and 1.6 m, and the central magnetic field was designed as 4.5/5.0 T. • The different maintenance ports design have little impact on the design of TF and CS coils’ design, but have certain impact on the PF coils’ design. -- Abstract: CFETR which stands for “China Fusion Engineering Test Reactor” is a new tokamak device. Its magnet system includes the Toroidal Field (TF) winding, Center solenoid winding (CS) and Poloidal Field (PF) winding. The main goal of the project is to build a fusion engineering Tokamak reactor with its fusion power is 50–200 MW and should be self-sufficiency by blanket. In order to ensure the maintenance ports design and maintenance method, this article discussed the concept design of the magnet system based on different maintenance port cases. The paper detailed studied the magnet system of CFETR including the electromagnetic analysis and parameters for TF (CS)PF. Besides, the volt-seconds of ohmic field are presented as detailed as possible in this paper. In addition, the calculations and optimizations of equilibrium field which should guarantee the plasma discharge of single null shape is carried out. The design work reported here illustrates that the present maintenance ports will not have a great impact on the design of the magnet system. The concept design of the magnet system can meet the requirement of the physical target

Automating expert role to determine design concept in Kansei Engineering

Science.gov (United States)

Lokman, Anitawati Mohd; Haron, Mohammad Bakri Che; Abidin, Siti Zaleha Zainal; Khalid, Noor Elaiza Abd

2016-02-01

Affect has become imperative in product quality. In affective design field, Kansei Engineering (KE) has been recognized as a technology that enables discovery of consumer's emotion and formulation of guide to design products that win consumers in the competitive market. Albeit powerful technology, there is no rule of thumb in its analysis and interpretation process. KE expertise is required to determine sets of related Kansei and the significant concept of emotion. Many research endeavors become handicapped with the limited number of available and accessible KE experts. This work is performed to simulate the role of experts with the use of Natphoric algorithm thus providing sound solution to the complexity and flexibility in KE. The algorithm is designed to learn the process by implementing training datasets taken from previous KE research works. A framework for automated KE is then designed to realize the development of automated KE system. A comparative analysis is performed to determine feasibility of the developed prototype to automate the process. The result shows that the significant Kansei is determined by manual KE implementation and the automated process is highly similar. KE research advocates will benefit this system to automatically determine significant design concepts.

Piping benchmark problems for the Westinghouse AP600 Standardized Plant

International Nuclear Information System (INIS)

Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

1997-01-01

To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for the Westinghouse AP600 Standardized Plant, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the AP600 standard design. It will be required that the combined license licensees demonstrate that their solutions to these problems are in agreement with the benchmark problem set

76 FR 441 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2011-01-05

... Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and Model C4-605R...-622, B4-605R, B4-622R, F4-605R, F4-622R, and C4-605R Variant F airplanes, certificated in any category...

76 FR 28914 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2011-05-19

... B4-600, B4-600R, and F4-600R Series Airplanes, and Model C4-605R Variant F Airplanes (Collectively... B4-605R and B4-622R airplanes; A300 F4-605R and F4-622R airplanes; A300 C4-605R Variant F airplanes...

75 FR 76926 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2010-12-10

... Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and Model C4-605R... airplanes; Model A300 B4-605R and B4-622R airplanes; Model A300 F4-605R and F4-622R airplanes; Model A300 C4...

76 FR 42029 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2011-07-18

... Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and Model A300 C4... A300 B4-605R and B4-622R airplanes, Model A300 F4-605R and F4-622R airplanes, and Model A300 C4- [[Page...

76 FR 38069 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2011-06-29

... B4-600, B4-600R, and F4-600R Series Airplanes, and Model C4-605R Variant F Airplanes (Collectively...-603, B4-620, B4-622, B4-605R, B4-622R, F4-605R, F4-622R, and C4-605R Variant F airplanes; and Model...

75 FR 27956 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2010-05-19

... B4-600, B4-600R, and F4-600R Series Airplanes, and Model C4-605R Variant F Airplanes (Collectively...-203, and B4-203 airplanes; Model A300 B4-601, B4- 603, B4-620, B4-622, B4-605R, B4-622R, F4-605R, F4...

76 FR 39248 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2011-07-06

... Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and Model C4-605R... B4-605R and B4-622R airplanes; Model A300 F4-605R and F4-622R airplanes; Model A300 C4-605R Variant F...

NASA Advanced Concepts Office, Earth-To-Orbit Team Design Process and Tools

Science.gov (United States)

Waters, Eric D.; Garcia, Jessica; Threet, Grady E., Jr.; Phillips, Alan

2013-01-01

The Earth-to-Orbit Team (ETO) of the Advanced Concepts Office (ACO) at NASA Marshall Space Flight Center (MSFC) is considered the pre-eminent "go-to" group for pre-phase A and phase A concept definition. Over the past several years the ETO team has evaluated thousands of launch vehicle concept variations for a significant number of studies including agency-wide efforts such as the Exploration Systems Architecture Study (ESAS), Constellation, Heavy Lift Launch Vehicle (HLLV), Augustine Report, Heavy Lift Propulsion Technology (HLPT), Human Exploration Framework Team (HEFT), and Space Launch System (SLS). The ACO ETO Team is called upon to address many needs in NASA's design community; some of these are defining extremely large trade-spaces, evaluating advanced technology concepts which have not been addressed by a large majority of the aerospace community, and the rapid turn-around of highly time critical actions. It is the time critical actions, those often limited by schedule or little advanced warning, that have forced the five member ETO team to develop a design process robust enough to handle their current output level in order to meet their customer's needs. Based on the number of vehicle concepts evaluated over the past year this output level averages to four completed vehicle concepts per day. Each of these completed vehicle concepts includes a full mass breakdown of the vehicle to a tertiary level of subsystem components and a vehicle trajectory analysis to determine optimized payload delivery to specified orbital parameters, flight environments, and delta v capability. A structural analysis of the vehicle to determine flight loads based on the trajectory output, material properties, and geometry of the concept is also performed. Due to working in this fast-paced and sometimes rapidly changing environment, the ETO Team has developed a finely tuned process to maximize their delivery capabilities. The objective of this paper is to describe the interfaces

Optimization and limitations of known DEMO divertor concepts

Energy Technology Data Exchange (ETDEWEB)

Reiser, Jens, E-mail: Jens.Reiser@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, 76021 Karlsruhe (Germany); Rieth, Michael [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, 76021 Karlsruhe (Germany)

2012-08-15

Highlights: Black-Right-Pointing-Pointer Limitations of the materials. Black-Right-Pointing-Pointer Improved H{sub 2}O cooled divertor. Black-Right-Pointing-Pointer Improved He cooled divertor. - Abstract: In this work we will introduce and discuss improvements for two types of DEMO divertors based on known designs: (i) gas cooled designs and (ii) liquid coolant concepts. In a first step, the advantages and disadvantages of gas cooling as well as the necessity of a jet impingement to increase the heat transfer coefficients will be discussed. Further discussion deals with the pros and cons of liquid coolant concepts, like for example, liquid metal or water cooling. Thereafter, we will present two rather contrary DEMO divertor concepts which are based on today's knowledge on refractory materials science, fabrication and joining technology. The first improved concept uses water flowing through steel pipes, typically made of Eurofer steel. It is well known that using Eurofer at low temperatures is critical due to its severe embrittlement under neutron irradiation. Here we make a proposal how it could be possible to use the Eurofer steel anyway: the solution could consist in a limited operation period followed by an annealing cycle at 550 Degree-Sign C for a few hours during any maintenance shut down phases. The second design is based on the known helium cooling concept using jet impingement. Drawbacks of the actual He-cooled divertor design are small scale parts as well as the necessary high helium inlet temperature of about 600-800 Degree-Sign C which leads to the question: How can we deal with such high helium temperatures? This paper shows a solution for large scale components as well as a new thermal management for the helium outlet gas that we call 'cooling of the coolant'. Both concepts are discussed in terms of materials selection due to material limits and joining technology with a special focus on the material issue using already existing and

Optimization and limitations of known DEMO divertor concepts

International Nuclear Information System (INIS)

Reiser, Jens; Rieth, Michael

2012-01-01

Highlights: ► Limitations of the materials. ► Improved H 2 O cooled divertor. ► Improved He cooled divertor. - Abstract: In this work we will introduce and discuss improvements for two types of DEMO divertors based on known designs: (i) gas cooled designs and (ii) liquid coolant concepts. In a first step, the advantages and disadvantages of gas cooling as well as the necessity of a jet impingement to increase the heat transfer coefficients will be discussed. Further discussion deals with the pros and cons of liquid coolant concepts, like for example, liquid metal or water cooling. Thereafter, we will present two rather contrary DEMO divertor concepts which are based on today's knowledge on refractory materials science, fabrication and joining technology. The first improved concept uses water flowing through steel pipes, typically made of Eurofer steel. It is well known that using Eurofer at low temperatures is critical due to its severe embrittlement under neutron irradiation. Here we make a proposal how it could be possible to use the Eurofer steel anyway: the solution could consist in a limited operation period followed by an annealing cycle at 550 °C for a few hours during any maintenance shut down phases. The second design is based on the known helium cooling concept using jet impingement. Drawbacks of the actual He-cooled divertor design are small scale parts as well as the necessary high helium inlet temperature of about 600–800 °C which leads to the question: How can we deal with such high helium temperatures? This paper shows a solution for large scale components as well as a new thermal management for the helium outlet gas that we call ‘cooling of the coolant’. Both concepts are discussed in terms of materials selection due to material limits and joining technology with a special focus on the material issue using already existing and available materials.

Concept Design of the Payload Handling Manipulator System. [space shuttle orbiters

Science.gov (United States)

1975-01-01

The design, requirements, and interface definition of a remote manipulator system developed to handle orbiter payloads are presented. End effector design, control system concepts, and man-machine engineering are considered along with crew station requirements and closed circuit television system performance requirements.

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

A 600 MWe advanced PWR for the 1990's

International Nuclear Information System (INIS)

Lemon, J.E.; Malloy, J.D.; Allen, R.E.

1987-01-01

The Babcock and Wilcox Company (B and W) and United Engineers and Constructors (UE and C) have prepared a conceptual design of an advanced 600 MWe Presurized Water Nuclear Power Plant. This design utilizes the large body of design and operating experience on PWRs in the U.S. and abroad and incorporates improvements emphasizing simplicity, safety, licensability, ease of construction, operability, reliability and maintainability. Cost and schedule estimates based on U.S. utility experience indicate that this plant design should be competitive with alternate options

Alternate design concept for the SSC dipole magnet cryogenic support post

International Nuclear Information System (INIS)

Lipski, A.; Nicol, T.H.; Richardson, R.

1991-03-01

New materials and developments in the field of advanced composites have created the opportunity to take a fresh look into the design of the cryogenic supports for SSC collider dipole cryostats. Although the present reentrant post design meets the structural and thermal requirements, its assembly requires precision and proficiency. The objective of the proposed alternate concept is to reduce the overall cost of the support post by means of simplifying and optimizing its component design and assembly process. The present shrink fitted tube assembly may potentially be replaced by injection molded parts. New resin systems with lower thermal conductivity and high strength properties enable the utilization of automated production techniques such as injection molding and filament winding. This paper will provide analysis and design information for the alternate support post concept and compare its test performance and cost to the present support post. 3 refs., 12 figs., 4 tabs

21 CFR 73.600 - Turmeric.

Science.gov (United States)

2010-04-01

... ADDITIVES EXEMPT FROM CERTIFICATION Foods § 73.600 Turmeric. (a) Identity. (1) The color additive turmeric is the ground rhizome of Curcuma longa L. The definition of turmeric in this paragraph is for the... 21 Food and Drugs 1 2010-04-01 2010-04-01 false Turmeric. 73.600 Section 73.600 Food and Drugs...

A new mix design concept for earth-moist concrete: A theoretical and experimental study

NARCIS (Netherlands)

Hüsken, Götz; Brouwers, Jos

2008-01-01

This paper addresses experiments on earth-moist concrete (EMC) based on the ideas of a new mix design concept. First, a brief introduction into particle packing and relevant packing theories is given. Based on packing theories for geometric packing, a new concept for the mix design of earth-moist

76 FR 27242 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2011-05-11

... Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and Model C4-605R... applies to all Airbus Model A300 B4-601, B4-603, B4- 620, B4-622, B4-605R, B4-622R, F4-605R, F4-622R, and...

Thermal design, analysis and comparison on three concepts of space solar power satellite

Science.gov (United States)

Yang, Chen; Hou, Xinbin; Wang, Li

2017-08-01

Space solar power satellites (SSPS) have been widely studied as systems for collecting solar energy in space and transmitting it wirelessly to earth. A previously designed planar SSPS concept collects solar power in two huge arrays and then transmits it through one side of the power-conduction joint to the antenna. However, the system's one group of power-conduction joints may induce a single point of failure. As an SSPS concept, the module symmetrical concentrator (MSC) architecture has many advantages. This architecture can help avoid the need for a large, potentially failure-prone conductive rotating joint and limit wiring mass. However, the thermal control system has severely restricted the rapid development of MSC, especially in the sandwich module. Because of the synchronous existence of five suns concentration and solar external heat flux, the sandwich module will have a very high temperature, which will surpass the permissible temperature of the solar cells. Recently, an alternate multi-rotary joints (MR) SSPS concept was designed by the China Academy of Space Technology (CAST). This system has multiple joints to avoid the problem of a single point of failure. Meanwhile, this concept has another advantage for reducing the high power and heat removal in joints. It is well known to us that, because of the huge external flux in SSPS, the thermal management sub-system is an important component that cannot be neglected. Based on the three SSPS concepts, this study investigated the thermal design and analysis of a 1-km, gigawatt-level transmitting antenna in SSPS. This study compares the thermal management sub-systems of power-conduction joints in planar and MR SSPS. Moreover, the study considers three classic thermal control architectures of the MSC's sandwich module: tile, step, and separation. The study also presents an elaborate parameter design, analysis and discussion of step architecture. Finally, the results show the thermal characteristics of each SSPS

Main results of BN-600 reactor stress-strain state investigations

International Nuclear Information System (INIS)

Panov, V.A.

1983-01-01

The development of BN-600 fast reactor plant needed the solution of a series of complex engineering problems including ones for confirming integrity of the most vital structural components. The particular attention was given to the main vessel since reactor availability end safe operation of the plant as a whole depend on vessel strength end integrity. The present report deals with the main results of theoretical and experimental investigations of the stress-strain state of BN-600 reactor vessel carried out during design, start-up and initial bringing the reactor to power

The Concept of Design: Are We off Track?

Science.gov (United States)

2012-06-15

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

Substantiation of physical concepts of fast reactors in Russia: experience and prospects

Energy Technology Data Exchange (ETDEWEB)

Alekseev, P.N. [Russian Research Center ' Kurchatov Institute' (RRC KI), 1, Kurchatov Sq., Moscow, 123182 (Russian Federation); Vasiliev, B.A. [Experimental Design Bureau of Machine Building (OKBM) 15, Burnakovskiy Pr., N. Novgorod, 603074 (Russian Federation); Kormilitsyn, M.V. [State Scientific Center of Russian Federation - Research Institute of Atomic Reactors (NIIAR) Dimitrovgrad-10, Ulianovsk Reg., 433510 (Russian Federation); Lopatkin, A.V. [N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) 2/8, M. Krasnoselskaya Str., Moscow, 107140 (Russian Federation); Seleznev, E.F. [All-Russian Research Institute for Nuclear Power Plant Operation (VNIIAES) 25, Ferganskaya, Moscow, 109507 (Russian Federation); Khomyakov, Yu.S.; Tsybulia, A.M. [State Scientific Center of the Russian Federation - A. I. Leypunsky Institute for Physics and Power Engineering (SSC RF- IPPE) 1, Bondarenko Sq., Obninsk, Kaluga Reg., 249033 (Russian Federation); Tocheny, L.V. [International Science and Technology Center (ISTC) 32-34 Krasnoproletarskaya Ulitsa, Moscow, 127473 (Russian Federation)

2008-07-01

The fast reactor concept in Russia has accumulated unique experience, since its advent in the 1950's and up to the present, from the creation of the first experimental installation BR-1, experimental reactors BR-5 and BOR-60, the pilot industrial reactors BN-350 in Kazakhstan and up to the BN-600 at Beloyarsk Atomic Power Station. Investigations on the first experimental installations BR-1 and BR-5/-10 proved the propriety of the idea that it is possible to create nuclear reactors that can produce more nuclear fuel than they consume, i.e. the idea of breeding. The architecture of such reactors was also designed, producing a current leader among fast reactors with sodium coolant and oxide uranium-plutonium fuel. Operational experience of BOR-60, BN-350 and, particularly, BN-600 confirmed the engineering and technical feasibility of the concept of fast reactors, the possibility for its realization both for power production and for certain other purposes as well, such as desalinisation of sea water (BN-350) and for radionuclide production (BN-350, BN-600), and it enabled the development and verification of different models, computer methods and codes. The paper presents a review of experience in the creation of plants with fast reactors, scientific research on these installations, principal results, the current status of experimental data analysis, and prospective directions in the development of fast reactors and the corresponding experimental basis in Russia. (authors)

Preliminary Assessment of Transient of Over Power Accident for DSFR-600

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Bae, Moohoon; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

TRACE code was selected as one of candidates for audit code, so sodium properties and heat transfer model in the code was verified first. On the basis of MARS-LMR code input, DSFR-600 TRACE model was developed and applied to PHTS tube rupture case, one of design base events (DBE) of DSFR-600. In this study, Transients of Over Power (TOP) event is assessed using TRACE code as one another case of DBEs of DSFR-600 for preparation of audit calculation of PGSFR.One of the design base events, transients of over power of Demonstration Sodium cooled Fast Reactor was simulated using TRACE code. Predicted fuel temperature showed that the peak fuel temperature occurs when the reactor scrammed and predicted temperature was similar to the MARS-LMRs assessment by KAERI. In this study, it is found that the second peak of fuel temperature is influenced by the inventory of steam generator and the natural circulation characteristic of the reactor vessel pool. Pre-calculation of the unprotected transients of over power with conservative reactivity assumption showed that this assumption is conservative in design base even assessment. However the method of measurement and applying the core radial, fuel and control rod axial expansion reactivity feedback is crucial in BDBE assessment of SFR.

MARS, 600 MWth NUCLEAR POWER PLANT

International Nuclear Information System (INIS)

Cumo, M.; Naviglio, A.; Sorabella, L.

2004-01-01

MARS (Multipurpose Advanced Reactor, inherently Safe) is a 600 MWth, single loop, pressurized light water reactor (PWR), developed at the Dept. of Nuclear Engineering and Energy Conversion of the University of Rome ''La Sapienza''. The design was focused to a multipurpose reactor to be used in high population density areas also for industrial heat production and, in particular, for water desalting. Using the well-proven technology and the operation experience of PWRs, the project introduces a lot of innovative features hugely improving the safety performance while keeping the cost of KWh competitive with traditional large power plants. Extensive use of passive safety, in depth plant simplification and decommissioning oriented design were the guidelines along the design development. The latest development in the plant design, in the decommissioning aspects and in the experimental activities supporting the project are shown in this paper

76 FR 59008 - Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and...

Science.gov (United States)

2011-09-23

... Airworthiness Directives; Airbus Model A300 B4-600, B4-600R, and F4-600R Series Airplanes, and Model C4-605R... to Airbus Model A300 B4-601, B4-603, B4-620, B4-622, B4-605R, B4-622R, F4-605R, F4-622R, and C4-605R...

75 FR 28480 - Airworthiness Directives; Airbus Model A300 Series Airplanes; Model A300 B4-600, B4-600R, F4-600R...

Science.gov (United States)

2010-05-21

... Airworthiness Directives; Airbus Model A300 Series Airplanes; Model A300 B4-600, B4-600R, F4-600R Series..., B4-622, B4- 605R, B4-622R, F4-605R, F4-622R, and C4-605R Variant F airplanes; and Model A310-203...