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Sample records for salt systems code

  1. SALT [System Analysis Language Translater]: A steady state and dynamic systems code

    International Nuclear Information System (INIS)

    Berry, G.; Geyer, H.

    1983-01-01

    SALT (System Analysis Language Translater) is a lumped parameter approach to system analysis which is totally modular. The modules are all precompiled and only the main program, which is generated by SALT, needs to be compiled for each unique system configuration. This is a departure from other lumped parameter codes where all models are written by MACROS and then compiled for each unique configuration, usually after all of the models are lumped together and sorted to eliminate undetermined variables. The SALT code contains a robust and sophisticated steady-sate finder (non-linear equation solver), optimization capability and enhanced GEAR integration scheme which makes use of sparsity and algebraic constraints. The SALT systems code has been used for various technologies. The code was originally developed for open-cycle magnetohydrodynamic (MHD) systems. It was easily extended to liquid metal MHD systems by simply adding the appropriate models and property libraries. Similarly, the model and property libraries were expanded to handle fuel cell systems, flue gas desulfurization systems, combined cycle gasification systems, fluidized bed combustion systems, ocean thermal energy conversion systems, geothermal systems, nuclear systems, and conventional coal-fired power plants. Obviously, the SALT systems code is extremely flexible to be able to handle all of these diverse systems. At present, the dynamic option has only been used for LMFBR nuclear power plants and geothermal power plants. However, it can easily be extended to other systems and can be used for analyzing control problems. 12 refs

  2. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

    2016-08-15

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  3. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    International Nuclear Information System (INIS)

    Shi, Chengbin; Cheng, Maosong; Liu, Guimin

    2016-01-01

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  4. SALT4: a two-dimensional displacement discontinuity code for thermomechanical analysis in bedded salt deposits

    International Nuclear Information System (INIS)

    1983-04-01

    SALT4 is a two-dimensional analytical/displacement-discontinuity code designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. This code was developed by the University of Minnesota. This documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of the computer code, SALT4. The SALT4 code takes into account: (1) viscoelastic behavior in the pillars adjacent to excavations; (2) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (2) excavation sequence. Major advantages of the SALT4 code are: (1) computational efficiency; (2) the small amount of input data required; and (3) a creep law consistent with laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of SALT4, i.e., temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT4 code can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermal and thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT4 can also be used to verify fully numerical codes. This is similar to the use of analytic solutions for code verification. Although SALT4 was designed for analysis of bedded salt, it is also applicable to crystalline rock if the creep calculation is suppressed. In Section 1.5 of this document the code custodianship and control is described along with the status of verification, validation and peer review of this report

  5. User's manual and guide to SALT3 and SALT4: two-dimensional computer codes for analysis of test-scale underground excavations for the disposal of radioactive waste in bedded salt deposits

    International Nuclear Information System (INIS)

    Lindner, E.N.; St John, C.M.; Hart, R.D.

    1984-02-01

    SALT3 and SALT4 are two-dimensional analytical/displacement-discontinuity codes designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. These codes were developed by the University of Minnesota for the Office of Nuclear Waste Isolation in 1979. The present documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of these computer codes. The SALT3 and SALT4 codes can simulate: (a) viscoelastic behavior in pillars adjacent to excavations; (b) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (c) excavation sequence. Major advantages of these codes are: (a) computational efficiency; (b) the small amount of input data required; and (c) a creep law based on laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of the codes, i.e., the homogeneous elastic half-space and temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT3 and SALT4 codes can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT3 is a refinement of an earlier code, SALT, and includes a fully anelastic creep model and thermal stress routine. SALT4 is a later version, and incorporates a revised creep model which is strain-hardening

  6. Development of a three dimension multi-physics code for molten salt fast reactor

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2014-01-01

    Molten Salt Reactor (MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum (GIF). The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors. In the present paper: a new coupling model is presented that physically describes the inherent relations between the neutron flux, the delayed neutron precursor, the heat transfer and the turbulent flow. Based on the model, integrating nuclear data processing, CAD modeling, structured and unstructured mesh technology, data analysis and visualization application, a three dimension steady state simulation code system (MSR3DS) for the can-type molten salt fast reactor is developed and validated. In order to demonstrate the ability of the code, the three dimension distributions of the velocity, the neutron flux, the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter (MOSART) using this code. The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor. Furthermore, the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion. (authors)

  7. Transient Analyses for a Molten Salt Transmutation Reactor Using the Extended SIMMER-III Code

    International Nuclear Information System (INIS)

    Wang, Shisheng; Rineiski, Andrei; Maschek, Werner; Ignatiev, Victor

    2006-01-01

    Recent developments extending the capabilities of the SIMMER-III code for the dealing with transient and accidents in Molten Salt Reactors (MSRs) are presented. These extensions refer to the movable precursor modeling within the space-time dependent neutronics framework of SIMMER-III, to the molten salt flow modeling, and to new equations of state for various salts. An important new SIMMER-III feature is that the space-time distribution of the various precursor families with different decay constants can be computed and took into account in neutron/reactivity balance calculations and, if necessary, visualized. The system is coded and tested for a molten salt transmuter. This new feature is also of interest in core disruptive accidents of fast reactors when the core melts and the molten fuel is redistributed. (authors)

  8. Development of a safety analysis code for molten salt reactors

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui

    2009-01-01

    The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.

  9. Development of system analysis code for pyrochemical process using molten salt electrorefining

    International Nuclear Information System (INIS)

    Tozawa, K.; Matsumoto, T.; Kakehi, I.

    2000-04-01

    This report describes accomplishment of development of a cathode processor calculation code to simulate the mass and heat transfer phenomena with the distillation process and development of an analytical model for cooling behavior of the pyrochemical process cell on personal computers. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. The cathode processor calculation code with distillation process was developed. A code validation calculation has been conducted on the basic of the benchmark problem for natural convection in a square cavity. Results by using the present code agreed well for the velocity-temperature fields, the maximum velocity and its location with the benchmark solution published in a paper. The functions have been added to advance the reality in simulation and to increase the efficiency in utilization. The test run has been conducted using the code with the above modification for an axisymmetric enclosed vessel simulating a cathode processor, and the capability of the distillation process simulation with the code has been confirmed. An analytical model for cooling behavior of the pyrochemical process cell was developed. The analytical model was selected by comparing benchmark analysis with detailed analysis on engineering workstation. Flow and temperature distributions were confirmed by the result of steady state analysis. In the result of transient cooling analysis, an initial transient peak of temperature occurred at balanced heat condition in the steady-state analysis. Final gas temperature distribution was dependent on gas circulation flow in transient condition. Then there were different final gas temperature distributions on the basis of the result of steady-state analysis. This phenomenon has a potential for it's own metastable condition. Therefore it was necessary to design gas cooling flow pattern without cooling gas circulation

  10. Benchmark Simulation of Natural Circulation Cooling System with Salt Working Fluid Using SAM

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, K. K.; Scarlat, R. O.; Hu, R.

    2017-09-03

    Liquid salt-cooled reactors, such as the Fluoride Salt-Cooled High-Temperature Reactor (FHR), offer passive decay heat removal through natural circulation using Direct Reactor Auxiliary Cooling System (DRACS) loops. The behavior of such systems should be well-understood through performance analysis. The advanced system thermal-hydraulics tool System Analysis Module (SAM) from Argonne National Laboratory has been selected for this purpose. The work presented here is part of a larger study in which SAM modeling capabilities are being enhanced for the system analyses of FHR or Molten Salt Reactors (MSR). Liquid salt thermophysical properties have been implemented in SAM, as well as properties of Dowtherm A, which is used as a simulant fluid for scaled experiments, for future code validation studies. Additional physics modules to represent phenomena specific to salt-cooled reactors, such as freezing of coolant, are being implemented in SAM. This study presents a useful first benchmark for the applicability of SAM to liquid salt-cooled reactors: it provides steady-state and transient comparisons for a salt reactor system. A RELAP5-3D model of the Mark-1 Pebble-Bed FHR (Mk1 PB-FHR), and in particular its DRACS loop for emergency heat removal, provides steady state and transient results for flow rates and temperatures in the system that are used here for code-to-code comparison with SAM. The transient studied is a loss of forced circulation with SCRAM event. To the knowledge of the authors, this is the first application of SAM to FHR or any other molten salt reactors. While building these models in SAM, any gaps in the code’s capability to simulate such systems are identified and addressed immediately, or listed as future improvements to the code.

  11. COSA II Further benchmark exercises to compare geomechanical computer codes for salt

    International Nuclear Information System (INIS)

    Lowe, M.J.S.; Knowles, N.C.

    1989-01-01

    Project COSA (COmputer COdes COmparison for SAlt) was a benchmarking exercise involving the numerical modelling of the geomechanical behaviour of heated rock salt. Its main objective was to assess the current European capability to predict the geomechanical behaviour of salt, in the context of the disposal of heat-producing radioactive waste in salt formations. Twelve organisations participated in the exercise in which their solutions to a number of benchmark problems were compared. The project was organised in two distinct phases: The first, from 1984-1986, concentrated on the verification of the computer codes. The second, from 1986-1988 progressed to validation, using three in-situ experiments at the Asse research facility in West Germany as a basis for comparison. This document reports the activities of the second phase of the project and presents the results, assessments and conclusions

  12. Improvements and validation of the transient analysis code MOREL for molten salt reactors

    International Nuclear Information System (INIS)

    Zhuang Kun; Zheng Youqi; Cao Liangzhi; Hu Tianliang; Wu Hongchun

    2017-01-01

    The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective. (author)

  13. The community project COSA: comparison of geo-mechanical computer codes for salt

    International Nuclear Information System (INIS)

    Lowe, M.J.S.; Knowles, N.C.

    1986-01-01

    Two benchmark problems related to waste disposal in salt were tackled by ten European organisations using twelve rock-mechanics finite element computer codes. The two problems represented increasing complexity with first a hypothetical verification and then the simulation of a laboratory experiment. The project allowed to ascertain a shapshot of the current combined expertise of European organisations in the modelling of salt behaviour

  14. Repository seal materials performance for a SALT Repository Project 5-year code/model development plan: Draft

    International Nuclear Information System (INIS)

    1987-06-01

    This document describes an integrated laboratory testing and model development effort for the seal system for a high-level nuclear waste repository in salt. The testing and modeling efforts are designed to determine seal material response in the repository environment, to provide models of seal system components for performance assessment, and to assist in the development of seal system designs. A code/model development and performance analysis program will be performed to predict the short- and long-term response of seal materials and seal components. The results from these analyses will be used to support the material testing activities on this contract and to support performance assessment activities that are conducted in other parts of the Salt Repository Project (SRP). 48 refs., 15 figs., 4 tabs

  15. Three-dimensional numerical investigation of a Molten Salt reactor concept with the code CFX-5.5

    International Nuclear Information System (INIS)

    Yamaji, B.; Csom, G.; Aszodi, A.

    2002-01-01

    Partitioning and transmutation of actinides and long-lived fission products is a promising option to extend the possibilities and enhance the environmentally acceptable capabilities of nuclear energy. Also the possible implementation of the thorium cycle is considered as a way to reduce the problem of energy resources in the future. For both objectives different molten salt reactor concepts were proposed mainly based on the Molten Salt Reactor Experiment of the Oak Ridge National Laboratory. Not only critical reactors but also accelerator-driven subcritical systems (ADSs) have advantages worth considering for those aims, especially those ones with liquid fuel, such as molten salts. By using liquid fuel which is the coolant medium, too, a basically different thermalhydraulic behavior is expected than in the case of solid fuel and water coolant. In this work our purpose is to present the possible use of Computational Fluid Dynamics (CFD) technology in molten salt thermal hydraulics. The simulations were performed with the three-dimensional code CFX-5.5.(author)

  16. Molten salt related extensions of the SIMMER-III code and its application for a burner reactor

    International Nuclear Information System (INIS)

    Wang Shisheng; Rineiski, Andrei; Maschek, Werner

    2006-01-01

    Molten salt reactors (MSRs) can be used as effective burners of plutonium (Pu) and minor actinides (MAs) from light water reactor (LWR) spent fuel. In this paper a study was made to examine the thermal hydraulic behaviour of the conceptual design of the molten salt advanced reactor transmuter (MOSART) [Ignatiev, V., Feynberg, O., Myasnikov, A., Zakirov, R., 2003a. Neutronic properties and possible fuel cycle of a molten salt transmuter. Proceedings of the 2003 ANS/ENS International Winter Meeting (GLOBAL 2003), Hyatt Regency, New Orleans, LA, USA 16-20 November 2003]. The molten salt fuel is a ternary NaF-LiF-BeF 2 system fuelled with ca. 1 mol% typical compositions of transuranium-trifluorides (PuF 3 , etc.) from light water reactor spent fuel. The MOSART reactor core does not contain graphite structure elements to guide the flow, so the neutron spectrum is rather hard in order to improve the burning performance. Without those structure elements in the core, the molten salt in core flows freely and the flow pattern could be potentially complicated and may affect significantly the fuel temperature distribution in the core. Therefore, some optimizations of the salt flow pattern may be needed. Here, the main attention has been paid to the fluid dynamic simulations of the MOSART core with the code SIMMER-III [Kondo, Sa., Morita, K., Tobita, Y., Shirakawa, K., 1992. SIMMER-III: an advanced computer program for LMFBR severe accident analysis. Proceedings of the ANP' 92, Tokyo, Japan; Kondo, Sa., Tobita, Y., Morita, K., Brear, D.J., Kamiyama, K., Yamano, H., Fujita, S., Maschek, W., Fischer, E.A., Kiefhaber, E., Buckel, G., Hesselschwerdt, E., Flad, M., Costa, P., Pigny, S., 1999. Current status and validation of the SIMMER-III LMFR safety analysis code. Proceedings of the ICONE-7, Tokyo, Japan], which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors for the thermo-hydraulic and neutronic models so as

  17. Numerical Analyses of a single-phase natural convection system for Molten Flibe using MARS-FLIBE code

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Sarah; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-10-15

    These advantages make the MSR attractive and to be one of the six candidates for the Generation IV Reactor. Therefore, the researches related to the MSR are being conducted. To analyze the molten salt-cooled systems in the laboratory, this study generated the properties of molten salt using MARS-LMR. In this research, the implemented salts were Flibe (LiF-BeF{sub 2}) in a molar mixture that is 66% LiF and 34% BeF{sub 2}, respectively. Table 1 indicates the comparison of thermal properties of various coolants in nuclear power plants. Molten salt was added to the MARS-LMR code to support the analysis of Flibe-cooled systems. The molten salt includes LiF-BeF{sub 2} in a molar mixture that is 66% LiF and 34% BeF{sub 2}, respectively. MARS-LMR code for liquid metals uses the soft sphere model based on Monte Carlo calculations for particles interacting with pair potentials. Although MARS was originally intended for a safety analysis of light water reactor, Flibe properties were newly added to this code as so-called MARS-FLIBE which is applicable for Flibe-cooled systems. By using this thermodynamic property table file, the thermal hydraulic systems of Flibe can be simulated for numerical and parametric studies. In this study, the natural convection phenomena in the rectangular natural convection loop and IVR-ERVC in APR 1400 were simulated. Through the simulations in Flibe-cooled systems, the temperature distribution and mass flowrate of Flibe can be calculated and the heat transfer coefficients of Flibe in natural convection loop will be calculated by adding the related heat transfer correlations in the MARS-FLIBE code. MARS-FLIBE code will be used to predict and design of Flibe-cooled systems.

  18. Studies on the molten salt reactor. Code development and neutronics analysis of MSRE-type design

    International Nuclear Information System (INIS)

    Zhuang Kun; Cao Liangzhi; Zheng Youqi; Wu Hongchun

    2015-01-01

    The molten salt reactor is characterized by its use of the fluid-fuel, which serves both as a fuel and as a coolant simultaneously. The position of delayed neutron precursors continuously changes both in the core and in the external loop due to the fuel circulation, and the fission products are extracted by an online fuel reprocessing unit, which all lead to the modeling methods for the conventional reactors using solid fuel not applicable. This study establishes suitable calculation models for the neutronics analysis of the molten salt reactor and develops a new code named MOREL based on the three-dimensional diffusion steady and transient calculations. Some numerical tests are chosen to verify the code and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactor. After verification, it is applied to analyze the characteristics of a typical molten salt reactor, including the steady characteristics, the influence of fuel circulation on the kinetic behaviors. Besides, the influence of online fuel reprocessing simulation is also examined. The results show that inherent safety is the character of the molten salt reactor from the aspect of reactivity feedback and the fuel circulation has great influence on the kinetic characteristics of molten salt reactor. (author)

  19. Validation of the TRACE code for the system dynamic simulations of the molten salt reactor experiment and the preliminary study on the dual fluid molten salt reactor

    International Nuclear Information System (INIS)

    He, Xun

    2016-01-01

    Molten Salt Reactor (MSR), which was confirmed as one of the six Generation IV reactor types by the GIF (Generation IV International Forum in 2008), recently draws a lot of attention all around the world. Due to the application of liquid fuels the MSR can be regarded as the most special one among those six GEN-IV reactor types in a sense. A unique advantage of using liquid nuclear fuel lies in that the core melting accident can be thoroughly eliminated. Besides, a molten salt reactor can have several fuel options, for instance, the fuel can be based on "2"3"5U, "2"3"2Th-"2"3"3U, "2"3"8U-"2"3"9Pu cycle or even the spent nuclear fuel (SNF), so the reactor can be operated as a breeder or as an actinides burner both with fast, thermal or epi-thermal neutron spectrum and hence, it has excellent features of the fuel sustainability and for the non-proliferation. Furthermore, the lower operating pressure not only means a lower risk of the explosion as well as the radioactive leakage but also implies that the reactor vessel and its components can be lightweight, thus lowering the cost of equipments. So far there is no commercial MSR being operated. However, the MSR concept and its technical validation dates back to the 1960s to 1970s, when the scientists and engineers from ORNL (Oak Ridge National Laboratory) in the United States managed to build and run the world's first civilian molten salt reactor called MSRE (Molten Salt Reactor Experiment). The MSRE was an experimental liquid-fueled reactor with 10 MW thermal output using "4LiF-BeF_2-ZrF_4-UF_4 as the fuel also as the coolant itself. The MSRE is usually taken as a very important reference case for many current researches to validate their codes and simulations. Without exception it works also as a benchmark for this thesis. The current thesis actually consists of two main parts. The first part is about the validation of the current code for the old MSRE concept, while the second one is about the demonstration of a new

  20. Validation of the TRACE code for the system dynamic simulations of the molten salt reactor experiment and the preliminary study on the dual fluid molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    He, Xun

    2016-06-14

    Molten Salt Reactor (MSR), which was confirmed as one of the six Generation IV reactor types by the GIF (Generation IV International Forum in 2008), recently draws a lot of attention all around the world. Due to the application of liquid fuels the MSR can be regarded as the most special one among those six GEN-IV reactor types in a sense. A unique advantage of using liquid nuclear fuel lies in that the core melting accident can be thoroughly eliminated. Besides, a molten salt reactor can have several fuel options, for instance, the fuel can be based on {sup 235}U, {sup 232}Th-{sup 233}U, {sup 238}U-{sup 239}Pu cycle or even the spent nuclear fuel (SNF), so the reactor can be operated as a breeder or as an actinides burner both with fast, thermal or epi-thermal neutron spectrum and hence, it has excellent features of the fuel sustainability and for the non-proliferation. Furthermore, the lower operating pressure not only means a lower risk of the explosion as well as the radioactive leakage but also implies that the reactor vessel and its components can be lightweight, thus lowering the cost of equipments. So far there is no commercial MSR being operated. However, the MSR concept and its technical validation dates back to the 1960s to 1970s, when the scientists and engineers from ORNL (Oak Ridge National Laboratory) in the United States managed to build and run the world's first civilian molten salt reactor called MSRE (Molten Salt Reactor Experiment). The MSRE was an experimental liquid-fueled reactor with 10 MW thermal output using {sup 4}LiF-BeF{sub 2}-ZrF{sub 4}-UF{sub 4} as the fuel also as the coolant itself. The MSRE is usually taken as a very important reference case for many current researches to validate their codes and simulations. Without exception it works also as a benchmark for this thesis. The current thesis actually consists of two main parts. The first part is about the validation of the current code for the old MSRE concept, while the second

  1. Molten salt thermal energy storage systems: salt selection

    Energy Technology Data Exchange (ETDEWEB)

    Maru, H.C.; Dullea, J.F.; Huang, V.S.

    1976-08-01

    A research program aimed at the development of a molten salt thermal energy storage system commenced in June 1976. This topical report describes Work performed under Task I: Salt Selection is described. A total of 31 inorganic salts and salt mixtures, including 9 alkali and alkaline earth carbonate mixtures, were evaluated for their suitability as heat-of-fusion thermal energy storage materials at temperatures of 850 to 1000/sup 0/F. Thermophysical properties, safety hazards, corrosion, and cost of these salts were compared on a common basis. We concluded that because alkali carbonate mixtures show high thermal conductivity, low volumetric expansion on melting, low corrosivity and good stability, they are attractive as heat-of-fusion storage materials in this temperature range. A 35 wt percent Li/sub 2/CO/sub 3/-65 wt percent K/sub 2/CO/sub 3/ (50 mole percent Li/sub 2/CO/sub 3/-50 mole percent K/sub 2/CO/sub 3/) mixture was selected as a model system for further experimental work. This is a eutectoid mixture having a heat of fusion of 148 Btu/lb (82 cal/g) that forms an equimolar compound, LiKCO/sub 3/. The Li/sub 2/CO/sub 3/-K/sub 2/CO/sub 3/ mixture is intended to serve as a model system to define heat transfer characteristics, potential problems, and to provide ''first-cut'' engineering data required for the prototype system. The cost of a thermal energy storage system containing this mixture cannot be predicted until system characteristics are better defined. However, our comparison of different salts indicated that alkali and alkaline earth chlorides may be more attractive from a salt cost point of view. The long-term corrosion characteristics and the effects of volume change on melting for the chlorides should be investigated to determine their overall suitability as a heat-of-fusion storage medium.

  2. SYSNET: A salt-site systems network model for scenario assessments

    International Nuclear Information System (INIS)

    Reeves, M.; Banda, R.S.

    1986-12-01

    This document contains a description of the initial version of the systems model SYSNET. This model is being developed to analyze potentially disruptive scenarios of salt repository systems. Currently the model features a general three-dimensional network topology and simulates the processes of flow, heat transport in rock, heat transport in fluid, brine transport, salt creep dissolution, and precipitation. Of necessity, the SYSNET Code uses relatively simple semi-analytic algorithms so that it may be implemented statistically. Uncertain parameters may be sampled with a compatible preprocessor and then analyzed statistically with a compatible postprocessor. When used in this fashion, SYSNET may be caused to calculate distributions of various performance measures and sensitivities of performance measures to uncertain parameters. The ultimate objective of the SYSNET development is to prioritize data needs by computing sensitivities relative to a particular performance measure, namely the 10,000-year cumulative release, and to evaluate repository systems for compliance with the US Environmental Protection Agency (EPA) Standard. 16 refs., 25 figs., 31 tabs

  3. Geomechanics of bedded salt

    International Nuclear Information System (INIS)

    Serata, S.; Milnor, S.W.

    1979-01-01

    Creep data from the literature search is reinterpreted by SGI, resulting in a better understanding of the temperature and stress state dependence of the octahedral creep rate and the octahedral shear strength. The concept of a transition strength between the elastic and the plastic states is in agreement with the data. The elastic and rheological properties of salt are described, and a set of constitutive equations is presented. The dependence of material properties on parameters such as temperature is considered. Findings on the permeability of salt are summarized, and the in-situ behavior of openings in bedded salt is described based on extensive engineering experience. A stress measuring system utilizing a finite element computer code is discussed. Geological factors affecting the stability of salt openings are considered, and the Stress Control Technique for designing stable openings in bedded salt formations is explained

  4. A Nucleus-localized Long Non-Coding RNA Enhances Drought and Salt Stress Tolerance

    KAUST Repository

    Qin, Tao

    2017-09-09

    Long non-coding RNAs (lncRNAs) affect gene expression through a wide range of mechanisms and are considered as important regulators in many essential biological processes. A large number of lncRNA transcripts have been predicted or identified in plants in recent years. However, the biological functions for most of them are still unknown. In this study, we identified an Arabidopsis thaliana lncRNA, Drought induced RNA (DRIR), as a novel positive regulator of plant response to drought and salt stress. DRIR was expressed at a low level under non-stress conditions but can be significantly activated by drought and salt stress as well as by abscisic acid (ABA) treatment. We identified a T-DNA insertion mutant, drirD, which had higher expression of the DRIR gene than the wild type plants. The drirD mutant exhibits increased tolerance to drought and salt stress. Overexpressing DRIR in Arabidopsis also increased tolerance to drought and salt stress of the transgenic plants. The drirD mutant and the overexpressing seedlings are more sensitive to ABA than the wild type in stomata closure and seedling growth. Genome-wide transcriptome analysis demonstrated that the expression of a large number of genes was altered in drirD and the overexpressing plants. These include genes involved in ABA signaling, water transport and other stress-relief processes. Our study reveals a mechanism whereby DRIR regulates plant response to abiotic stress by modulating the expression of a series of genes involved in stress response.

  5. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  6. Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

    International Nuclear Information System (INIS)

    Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

    2002-01-01

    A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

  7. Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES)

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo; Mitachi, Koshi

    2013-01-01

    The authors have been promoting nuclear energy technology based on thorium molten salt as Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES). This system is a combination of fission power reactor of Molten Salt Reactor (MSR), and Accelerator Molten Salt Breeder (AMSB) for production of fissile 233 U with connecting chemical processing facility. In this paper, concept of THORIMS-NES, advantages of thorium and molten salt recent MSR design results such as FUJI-U3 using 233 U fuel, FUJI-Pu, large sized super-FUJI, pilot plant miniFUJI, AMSB, and chemical processing facility are described. (author)

  8. Evaluation of temporary non-code repairs in safety class 3 piping systems

    International Nuclear Information System (INIS)

    Godha, P.C.; Kupinski, M.; Azevedo, N.F.

    1996-01-01

    Temporary non-ASME Code repairs in safety class 3 pipe and piping components are permissible during plant operation in accordance with Nuclear Regulatory Commission Generic Letter 90-05. However, regulatory acceptance of such repairs requires the licensee to undertake several timely actions. Consistent with the requirements of GL 90-05, this paper presents an overview of the detailed evaluation and relief request process. The technical criteria encompasses both ductile and brittle piping materials. It also lists appropriate evaluation methods that a utility engineer can select to perform a structural integrity assessment for design basis loading conditions to support the use of temporary non-Code repair for degraded piping components. Most use of temporary non-code repairs at a nuclear generating station is in the service water system which is an essential safety related system providing the ultimate heat sink for various plant systems. Depending on the plant siting, the service water system may use fresh water or salt water as the cooling medium. Various degradation mechanisms including general corrosion, erosion/corrosion, pitting, microbiological corrosion, galvanic corrosion, under-deposit corrosion or a combination thereof continually challenge the pressure boundary structural integrity. A good source for description of corrosion degradation in cooling water systems is provided in a cited reference

  9. Molten-salt reactor information system

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Cardwell, D.W.; Engel, J.R.

    1975-06-01

    The Molten-Salt Reactor Information System (MSRIS) is a computer-based file of abstracts of documents dealing with the technology of molten-salt reactors. The file is stored in the IBM-360 system at ORNL, and may be searched through the use of established interactive computer programs from remote terminals connected to the computer via telephone lines. The system currently contains 373 entries and is subject to updating and expansion as additional information is developed. The nature and general content of the data file, a general approach for obtaining information from it, and the manner in which material is added to the file are described. Appendixes provide the list of keywords currently in use, the subject categories under which information is filed, and simplified procedures for searching the file from remote terminals. (U.S.)

  10. Symbiotic molten-salt systems coupled with accelerator molten-salt breeder (AMSB) or inertial-confined fusion hybrid molten-salt breeder (IHMSB) and their comparison

    International Nuclear Information System (INIS)

    Furukawa, K.

    1984-01-01

    Two types of breeder systems are proposed. One is the combined system of Accelerator Molten-Salt Breeder (AMSB) and Molten-Salt Converter Reactor (MSCR), and the other is the combined system of Inertial-confined Fusion Hybrid Molten-Salt Breeder (IHMSB) and modified MSCR. Both apply the molten-fluorides and have technically deep relations. AMSB would be much simpler and have already high technical feasibility. This will become economical the Th breeder system having a doubling time shorter than ten years and distributing any size of power stations MSCR. (orig.) [de

  11. Hydrophobic interaction chromatography in dual salt system increases protein binding capacity.

    Science.gov (United States)

    Senczuk, Anna M; Klinke, Ralph; Arakawa, Tsutomu; Vedantham, Ganesh; Yigzaw, Yinges

    2009-08-01

    Hydrophobic interaction chromatography (HIC) uses weakly hydrophobic resins and requires a salting-out salt to promote protein-resin interaction. The salting-out effects increase with protein and salt concentration. Dynamic binding capacity (DBC) is dependent on the binding constant, as well as on the flow characteristics during sample loading. DBC increases with the salt concentration but decreases with increasing flow rate. Dynamic and operational binding capacity have a major raw material cost/processing time impact on commercial scale production of monoclonal antibodies. In order to maximize DBC the highest salt concentration without causing precipitation is used. We report here a novel method to maintain protein solubility while increasing the DBC by using a combination of two salting-out salts (referred to as dual salt). In a series of experiments, we explored the dynamic capacity of a HIC resin (TosoBioscience Butyl 650M) with combinations of salts. Using a model antibody, we developed a system allowing us to increase the dynamic capacity up to twofold using the dual salt system over traditional, single salt system. We also investigated the application of this novel approach to several other proteins and salt combinations, and noted a similar protein solubility and DBC increase. The observed increase in DBC in the dual salt system was maintained at different linear flow rates and did not impact selectivity.

  12. Preliminary Study of Single-Phase Natural Circulation for Lab-scaled Molten Salt Application

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Yukyung; Kang, Sarah; Kim, In Guk; Seo, Seok Bin; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Park, Seong Dae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Advanced reactors such as MSR (FHR), VHTR and AHTR utilized molten salt as a coolant for efficiency and safety which has advantages in higher heat capacity, lower pumping power and scale compared to liquid metal. It becomes more necessary to study on the characteristics of molten salt. However, due to several characteristics such as high operating temperature, large-scale facility and preventing solidification, satisfying that condition for study has difficulties. Thus simulant fluid was used with scaling method for lab-scale experiment. Scaled experiment enables simulant fluid to simulate fluid mechanics and heat transfer behavior of molten salt on lower operating temperature and reduced scale. In this paper, as a proof test of the scaled experiment, simplified single-phase natural circulation loop was designed in a lab-scale and applied to the passive safety system in advanced reactor in which molten salt is considered as a major coolant of the system. For the application of the improved safety system, prototype was based on the primary loop of the test-scale DRACS, the main passive safety system in FHR, developed at the OSU. For preliminary experiment, single-phase natural circulation under low power was performed. DOWTHERM A and DOWTHERM RP were selected as simulant candidates. Then, study of feasibility with simulant was conducted based on the scaling law for heat transfer characteristics and geometric parameters. Additionally, simulation with MARS code and ANSYS-CFX with the same condition of natural circulation was carried out as verification. For the accurate code simulation, thermo-physical properties of DOWTHERM A and RP were developed and implemented into MARS code. In this study, single-phase natural circulation experiment was performed with simulant oil, DOWTHERM RP, based on the passive safety system of FHR. Feasibility of similarity experiment for molten salt with oil simulant was confirmed by scaling method. In addition, simulation with two

  13. Study of trans-uranian incineration in molten salt reactor

    International Nuclear Information System (INIS)

    Valade, M.

    2000-01-01

    With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

  14. Combined system of accelerator molten-salt breeder (AMSB) apd molten-salt converter reactor (MSCR)

    International Nuclear Information System (INIS)

    Furukawa, K.; Kato, Y.; Ohmichi, T.; Ohno, H.

    1983-01-01

    A design and research program is discUssed of the development of accelerator molten-salt breeder (AMSB) consisting of a proton accelerator and a molten fluoride target. The target simultaneously serves as a blanket for fissionable material prodUction. An addition of some amoUnt of fissile nuclides to a melt expands the AMSB potentialities as the fissionable material production increases and the energy generation also grows up to the level of self-provision. Besides the blanket salts may be used as nuclear fuel for molten-salt converter reactor (MSCR). The combined AM SB+MSCR system has better parameters as compared to other breeder reactors, molten-salt breeder reactors (MSBR) included

  15. Advances in molten salt electrochemistry towards future energy systems

    International Nuclear Information System (INIS)

    Ito, Yasuhiko

    2005-01-01

    This review article describes some selected novel molten salt electrochemical processes which have been created/developed by the author and his coworkers, with emphasis on the applications towards future energy systems. After showing a perspective of the applications of molten salt electrochemistry from the viewpoints of energy and environment, several selected topics are described in detail, which include nitride fuel cycle in a nuclear field, hydrogen energy system coupled with ammonia economy, thermally regenerative fuel cell systems, novel Si production process for solar cell and novel molten salt electrochemical processes for various energy and environment related functional materials including nitrides, rare earth-transition metal alloys, fine particles obtained by plasma-induced electrolysis, and carbon film. And finally, the author stresses again, the importance and potential of molten salt electrochemistry, and encourages young students, scientists and researchers to march in a procession hand in hand towards a bright future of molten salts. (author)

  16. SASSYS LMFBR systems code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

    1983-01-01

    The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

  17. Code development and analyses within the area of transmutation and safety

    International Nuclear Information System (INIS)

    Maschek, W.

    2002-01-01

    A strong code development is going on to meet various demands resulting from the development of dedicated reactors for transmutation and incineration. Code development is concerned with safety codes and general codes needed for assessing scenarios and transmutation strategies. Analyses concentrate on various ADS systems with solid and liquid molten salt fuels. Analyses deal with ADS Demo Plant (5th FP EU) and transmuters with advanced fuels

  18. Study on Utilization of Super Grade Plutonium in Molten Salt Reactor FUJI-U3 using CITATION Code

    Science.gov (United States)

    Wulandari, Cici; Waris, Abdul; Pramuditya, Syeilendra; Asril, Pramutadi AM; Novitrian

    2017-07-01

    FUJI-U3 type of Molten Salt Reactor (MSR) has a unique design since it consists of three core regions in order to avoid the replacement of graphite as moderator. MSR uses floride as a nuclear fuel salt with the most popular chemical composition is LiF-BeF2-ThF4-233UF4. ThF4 and 233UF4 are the fertile and fissile materials, respectively. On the other hand, LiF and BeF2 working as both fuel and heat transfer medium. In this study, the super grade plutonium will be utilized as substitution of 233U since plutonium is easier to be obtained compared to 233U as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2002 code with JENDL 3.2 as nuclear data library.

  19. ETR/ITER systems code

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

  20. ETR/ITER systems code

    International Nuclear Information System (INIS)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs

  1. Experimental and theoretical studies in Molten Salt Natural Circulation Loop (MSNCL)

    International Nuclear Information System (INIS)

    Srivastava, A.K.; Borgohain, A.; Jana, S.S.; Bagul, R.K.; Singh, R.R.; Maheshwari, N.K.; Belokar, D.G.; Vijayan, P.K.

    2014-12-01

    High Temperature Reactors (HTR) and solar thermal power plants use molten salt as a coolant, as it has low melting point and high boiling point, enabling us to operate the system at low pressure. Molten fluoride salt and molten nitrate salt are proposed as a candidate coolant for High Temperature Reactors (HTR) and solar power plant respectively. BARC is developing a 600 MWth pebble bed high temperature reactor, cooled by natural circulation of fluoride salt and capable of supplying process heat at 1000°C to facilitate hydrogen production by splitting water. Beside this, BARC is also developing a 2MWe solar power tower system using molten nitrate salt. With these requirements, a Molten Salt Natural Circulation Loop (MSNCL) has been designed, fabricated, installed and commissioned in Hall-7, BARC for thermal hydraulic, instrumentation development and material compatibility related studies. Steady state natural circulation experiments with molten nitrate salt (mixture of NaNO 3 and KNO 3 in 60:40 ratio) have been carried out in the loop at different power level. Various transients viz. startup of natural circulation, step power change, loss of heat sink and heater trip has also been studied in the loop. A well known steady state correlation given by Vijayan et. al. has been compared with experimental data. In-house developed code LeBENC has also been validated against all steady state and transient experimental results. The detailed description of MSNCL, steady state and transient experimental results and validation of in-house developed code LeBENC have been described in this report. (author)

  2. Development of High Throughput Salt Separation System with Integrated Liquid Salt Separation - Salt Distillation Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Sangwoon; Park, K. M.; Kim, J. G.; Jeong, J. H.; Lee, S. J.; Park, S. B.; Kim, S. S.

    2013-01-15

    The capacity of a salt distiller should be sufficiently large to reach the throughput of uranium electro-refining process. In this study, an assembly composing a liquid separation sieve and a distillation crucible was developed for the sequential operation of a liquid salt separation and a vacuum distillation in the same tower. The feasibility of the sequential salt separation was examined by the rotation test of the sieve-crucible assembly and sequential operation of a liquid salt separation and a vacuum distillation. The adhered salt in the uranium deposits was removed successfully. The salt content in the deposits was below 0.1 wt% after the sequential operation of the liquid salt separation - salt distillation. From the results of this study, it could be concluded that efficient salt separation can be realized by the sequential operation of liquid salt separation and vacuum distillation in one distillation tower since the operation procedures are simplified and no extra operation of cooling and reheating is necessary.

  3. Salt intake and the validity of a salt intake assessment system based on a 24-h dietary recall method in pregnant Japanese women.

    Science.gov (United States)

    Satoh, Michihiro; Tanno, Yumi; Hosaka, Miki; Metoki, Hirohito; Obara, Taku; Asayama, Kei; Hoshi, Kazuhiko; Suzuki, Masakuni; Mano, Nariyasu; Imai, Yutaka

    2015-01-01

    Information regarding salt intake in pregnant women in Japan is limited. An electronic system for the assessment of salt intake using a 24-h dietary recall method has been developed in Japan. The objectives of the present study were to investigate salt intake in pregnant women and to compare the salt intake estimated by the electronic salt intake assessment system with that measured by 24-h urinary salt excretion (24-hUNaCl). Data were collected on 24-hUNaCl and salt intake estimated by the salt intake assessment system for 35 pregnant Japanese women at approximately 20 weeks of gestation. The adjusted 24-hUNaCl (24-hUNaCl/[the number of urinations during the examination day--the number of missing urine collections] × the number of urinations during the examination day, g/day) was used as a standard. The mean adjusted 24-hUNaCl was 7.7 ± 2.5 g/day, and mean systolic/diastolic blood pressure values were 106.1 ± 8.6/62.8 ± 6.5 mmHg. The adjusted 24-hUNaCl was significantly correlated with the salt intake estimated by the salt intake assessment system (r = 0.47, p = 0.004). Bland-Altman analysis showed no significant mean difference (adjusted 24-hUNaCl--salt intake estimated by the assessment system = -0.36 g/day, p = 0.4) and no significant proportional bias (p = 0.1). These results suggest that pregnant women in Japan restrict their salt intake, at least when they are being examined for salt intake. They also suggest that repeated use of the described system may be useful in estimating salt intake in pregnant women.

  4. Molten Chloride Salts for Heat Transfer in Nuclear Systems

    Science.gov (United States)

    Ambrosek, James Wallace

    2011-12-01

    A forced convection loop was designed and constructed to examine the thermal-hydraulic performance of molten KCl-MgCl2 (68-32 at %) salt for use in nuclear co-generation facilities. As part of this research, methods for prediction of the thermo-physical properties of salt mixtures for selection of the coolant salt were studied. In addition, corrosion studies of 10 different alloys were exposed to the KCl-MgCl2 to determine a suitable construction material for the loop. Using experimental data found in literature for unary and binary salt systems, models were found, or developed to extrapolate the available experimental data to unstudied salt systems. These property models were then used to investigate the thermo-physical properties of the LINO3-NaNO3-KNO 3-Ca(NO3), system used in solar energy applications. Using these models, the density, viscosity, adiabatic compressibility, thermal conductivity, heat capacity, and melting temperatures of higher order systems can be approximated. These models may be applied to other molten salt systems. Coupons of 10 different alloys were exposed to the chloride salt for 100 hours at 850°C was undertaken to help determine with which alloy to construct the loop. Of the alloys exposed, Haynes 230 had the least amount of weight loss per area. Nickel and Hastelloy N performed best based on maximum depth of attack. Inconel 625 and 718 had a nearly uniform depletion of Cr from the surface of the sample. All other alloys tested had depletion of Cr along the grain boundaries. The Nb in Inconel 625 and 718 changed the way the Cr is depleted in these alloys. Grain-boundary engineering (GBE) of Incoloy 800H improved the corrosion resistance (weight loss and maximum depth of attack) by nearly 50% as compared to the as-received Incoloy 800H sample. A high temperature pump, thermal flow meter, and pressure differential device was designed, constructed and tested for use in the loop, The heat transfer of the molten chloride salt was found to

  5. Salt site performance assessment activities

    International Nuclear Information System (INIS)

    Kircher, J.F.; Gupta, S.K.

    1983-01-01

    During this year the first selection of the tools (codes) for performance assessments of potential salt sites have been tentatively selected and documented; the emphasis has shifted from code development to applications. During this period prior to detailed characterization of a salt site, the focus is on bounding calculations, sensitivity and with the data available. The development and application of improved methods for sensitivity and uncertainty analysis is a focus for the coming years activities and the subject of a following paper in these proceedings. Although the assessments to date are preliminary and based on admittedly scant data, the results indicate that suitable salt sites can be identified and repository subsystems designed which will meet the established criteria for protecting the health and safety of the public. 36 references, 5 figures, 2 tables

  6. Heat transfer investigation of molten salts under laminar and turbulent flow regimes

    International Nuclear Information System (INIS)

    Srivastava, A.K.; Vaidya, A.M.; Maheshwari, N.K.; Vijayan, P.K.

    2014-01-01

    High temperature reactor and solar thermal power plants use Molten Salt as a coolant, as it has low melting point and high boiling point, enabling us to operate the system at low pressure. Molten fluoride salt (eutectic mixture of LiF-NaF-KF) and molten nitrate salt (mixture of NaNO 3 and KNO 3 in 60:40 ratios by weight) are proposed as a candidate coolant for High Temperature Reactors (HTR) and solar power plant respectively. BARC is developing a 600 MWth pebble bed high temperature reactor, cooled by natural circulation of fluoride salt and capable of supplying process heat at 1000℃ to facilitate hydrogen production by splitting water. Beside this, BARC is also developing a 2MWe solar power tower system using molten nitrate salt as a primary coolant and storage medium. In order to design this, it is necessary to study the heat transfer characteristics of various molten salts. Most of the previous studies related to molten salts are based on the experimental works. These experiments essentially measured the physical properties of molten salts and their heat transfer characteristics. Ferri et al. introduced the property definitions for molten salts in the RELAP5 code to perform transient simulations at the ProvaCollettoriSolari (PCS) test facility. In this paper, a CFD analysis has been performed to study the heat transfer characteristics of molten fluoride salt and molten nitrate salt flowing in a circular pipe for various regimes of flow. Simulation is performed with the help of in-house developed CFD code, NAFA, acronym for Numerical Analysis of Flows in Axi-symmetric geometries. Uniform velocity and temperature distribution are set as the inlet boundary condition and pressure is employed at the outlet boundary condition. The inlet temperature for all simulation is set as 300℃ for nitrate salt and 500℃ for fluoride salt and the operating pressure is 1 atm in both the cases

  7. Salt Lakes of the African Rift System: A Valuable Research ...

    African Journals Online (AJOL)

    Salt Lakes of the African Rift System: A Valuable Research Opportunity for Insight into Nature's Concenrtated Multi-Electrolyte Science. JYN Philip, DMS Mosha. Abstract. The Tanzanian rift system salt lakes present significant cultural, ecological, recreational and economical values. Beyond the wealth of minerals, resources ...

  8. An Overview of Liquid Fluoride Salt Heat Transport Systems

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL

    2010-09-01

    Heat transport is central to all thermal-based forms of electricity generation. The ever increasing demand for higher thermal efficiency necessitates power generation cycles transitioning to progressively higher temperatures. Similarly, the desire to provide direct thermal coupling between heat sources and higher temperature chemical processes provides the underlying incentive to move toward higher temperature heat transfer loops. As the system temperature rises, the available materials and technology choices become progressively more limited. Superficially, fluoride salts at {approx}700 C resemble water at room temperature being optically transparent and having similar heat capacity, roughly three times the viscosity, and about twice the density. Fluoride salts are a leading candidate heat-transport material at high temperatures. Fluoride salts have been extensively used in specialized industrial processes for decades, yet they have not entered widespread deployment for general heat transport purposes. This report does not provide an exhaustive screening of potential heat transfer media and other high temperature liquids such as alkali metal carbonate eutectics or chloride salts may have economic or technological advantages. A particular advantage of fluoride salts is that the technology for their use is relatively mature as they were extensively studied during the 1940s-1970s as part of the U.S. Atomic Energy Commission's program to develop molten salt reactors (MSRs). However, the instrumentation, components, and practices for use of fluoride salts are not yet developed sufficiently for commercial implementation. This report provides an overview of the current understanding of the technologies involved in liquid salt heat transport (LSHT) along with providing references to the more detailed primary information resources. Much of the information presented here derives from the earlier MSR program. However, technology has evolved over the intervening years

  9. A simplified burnup calculation strategy with refueling in static molten salt reactor

    International Nuclear Information System (INIS)

    Srivastava, A.K.; Gupta, Anurag; Krishnani, P.D.

    2015-01-01

    Molten Salt Reactors, by nature can be refuelled and reprocessed online. Thus, a simulation methodology has to be developed which can consider online refueling and reprocessing aspect of the reactor. To cater such needs a simplified burnup calculation strategy to account for refueling and removal of molten salt fuel at any desired burnup has been identified in static molten salt reactor in batch mode as a first step of way forward. The features of in-house code ITRAN has been explored for such calculations. The code also enables us to estimate the reactivity introduced in the system due to removal of any number of considered nuclides at any burnup. The effect of refueling fresh fuel and removal of burned fuel has been studied in batch mode with in-house code ITRAN. The effect of refueling and burnup on change in reactivity per day has been analyzed. The analysis of removal of 233 Pa at a particular burnup has been carried out. The similar analysis has been performed for some other nuclides also. (author)

  10. Photobleachable Diazonium Salt-Phenolic Resin Two-Layer Resist System

    Science.gov (United States)

    Uchino, Shou-ichi; Iwayanagi, Takao; Hashimoto, Michiaki

    1988-01-01

    This article describes a new negative two-layer photoresist system formed by a simple, successive spin-coating method. An aqueous acetic acid solution of diazonium salt and poly(N-vinylpyrrolidone) is deposited so as to contact a phenolic resin film spin-coated on a silicon wafer. The diazonium salt diffuses into the phenolic resin layer after standing for several minutes. The residual solution on the phenolic resin film doped with diazonium salt is spun to form the diazonium salt-poly(N-vinylpyrrolidone) top layer. This forms a uniform two-layer resist without phase separation or striation. Upon UV exposure, the diazonium salt in the top layer bleaches to act as a CEL dye, while the diazonium salt in the bottom layer decomposes to cause insolubilization. Half μm line-and-space patterns are obtained with an i-line stepper using 4-diazo-N,N-dimethylaniline chloride zinc chloride double salt as the diazonium salt and a cresol novolac resin for the bottom polymer layer. The resist formation processes, insolubilization mechanism, and the resolution capability of the new two-layer resist are discussed.

  11. ESCADRE and ICARE code systems

    International Nuclear Information System (INIS)

    Reocreux, M.; Gauvain, J.

    1992-01-01

    The French sever accident code development program is following two parallel approaches: the first one is dealing with ''integral codes'' which are designed for giving immediate engineer answers, the second one is following a more mechanistic way in order to have the capability of detailed analysis of experiments, in order to get a better understanding of the scaling problem and reach a better confidence in plant calculations. In the first approach a complete system has been developed and is being used for practical cases: this is the ESCADRE system. In the second approach, a set of codes dealing first with primary circuit is being developed: a mechanistic core degradation code, ICARE, has been issued and is being coupled with the advanced thermalhydraulic code CATHARE. Fission product codes have been also coupled to CATHARE. The ''integral'' ESCADRE system and the mechanistic ICARE and associated codes are described. Their main characteristics are reviewed and the status of their development and assessment given. Future studies are finally discussed. 36 refs, 4 figs, 1 tab

  12. The CORSYS neutronics code system

    International Nuclear Information System (INIS)

    Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.

    1994-01-01

    The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs

  13. An Implementation of the Salt-Farm Monitoring System Using Wireless Sensor Network

    Science.gov (United States)

    Ju, Jonggil; Park, Ingon; Lee, Yongwoong; Cho, Jongsik; Cho, Hyunwook; Yoe, Hyun; Shin, Changsun

    In producing solar salt, natural environmental factors such as temperature, humidity, solar radiation, wind direction, wind speed and rain are essential elements which influence on the productivity and quality of salt. If we can manage the above mentioned environmental elements efficiently, we could achieve improved results in production of salt with good quality. To monitor and manage the natural environments, this paper suggests the Salt-Farm Monitoring System (SFMS) which is operated with renewable energy power. The system collects environmental factors directly from the environmental measure sensors and the sensor nodes. To implement a stand-alone system, we applied solar cell and wind generator to operate this system. Finally, we showed that the SFMS could monitor the salt-farm environments by using wireless sensor nodes and operate correctly without external power supply.

  14. Development of High-Temperature Transport System for Molten Salt in Pyroprocessing

    International Nuclear Information System (INIS)

    Lee, Sung Ho; Kim, In Tae; Park, Sung Bin

    2014-01-01

    The electrorefining process, which is a key process in pyroprocessing, is composed of two parts, electrorefining to deposit a uranium with a solid cathode and electrowinning to co-deposit TRU and RE with a liquid cadmium cathode (LCC). As the electrorefining operation proceedes, TRU and RE are accumulated in electrolyte LiCl-KCl salt, and after the electrorefining process, the molten salt used in an electrorefining reactor should by transported to the next process, the electrowinning process, to recover U/TRU/RE; Thus, a molten salt transfer system by suction is now being developed. An apparatus for suction transport experiments was designed and constructed for the development of high- temperature molten salt transport technology. Suction transport experiments were performed using LiC-KCl eutectic salt. The feasibility of pyro-reprocessing has been demonstrated through many laboratory-scale experiments. In pyroprocessing, a eutectic LiCl-KCl salt was used as a liquid elextrolyte for a recovery of actinides. However, reliable transport technologies for these high temperature liquids have not yet been developed. A preliminary study on high-temperature transport technology for molten salt by suction is now being carried out. In this study, three different salt transport technologies (gravity, suction pump, and centrifugal pump) were investigated to select the most suitable method for molten salt transport. An apparatus for suction transport experiments was designed and installed for the development of high-temperature molten salt transport technology. Basic preliminary suction transport experiments were carried out using the prepared LiC-KCl eutectic salt at 500 .deg. C to observe the transport behavior of LiCl-KCl molten salt. In addition, a PRIDE salt transport system was designed and installed for an engineering-scale salt transport demonstration. Several types of suction transport experiments using molten salt (LiCl-KCl eutectics) for the development of a high

  15. Development of High-Temperature Transport System for Molten Salt in Pyroprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sung Ho; Kim, In Tae; Park, Sung Bin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The electrorefining process, which is a key process in pyroprocessing, is composed of two parts, electrorefining to deposit a uranium with a solid cathode and electrowinning to co-deposit TRU and RE with a liquid cadmium cathode (LCC). As the electrorefining operation proceedes, TRU and RE are accumulated in electrolyte LiCl-KCl salt, and after the electrorefining process, the molten salt used in an electrorefining reactor should by transported to the next process, the electrowinning process, to recover U/TRU/RE; Thus, a molten salt transfer system by suction is now being developed. An apparatus for suction transport experiments was designed and constructed for the development of high- temperature molten salt transport technology. Suction transport experiments were performed using LiC-KCl eutectic salt. The feasibility of pyro-reprocessing has been demonstrated through many laboratory-scale experiments. In pyroprocessing, a eutectic LiCl-KCl salt was used as a liquid elextrolyte for a recovery of actinides. However, reliable transport technologies for these high temperature liquids have not yet been developed. A preliminary study on high-temperature transport technology for molten salt by suction is now being carried out. In this study, three different salt transport technologies (gravity, suction pump, and centrifugal pump) were investigated to select the most suitable method for molten salt transport. An apparatus for suction transport experiments was designed and installed for the development of high-temperature molten salt transport technology. Basic preliminary suction transport experiments were carried out using the prepared LiC-KCl eutectic salt at 500 .deg. C to observe the transport behavior of LiCl-KCl molten salt. In addition, a PRIDE salt transport system was designed and installed for an engineering-scale salt transport demonstration. Several types of suction transport experiments using molten salt (LiCl-KCl eutectics) for the development of a high

  16. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.

  17. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    International Nuclear Information System (INIS)

    Lee, Young Jin; Chung, Bub Dong

    2004-01-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures

  18. System Design Description for the TMAD Code

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    1995-01-01

    This document serves as the System Design Description (SDD) for the TMAD Code System, which includes the TMAD code and the LIBMAKR code. The SDD provides a detailed description of the theory behind the code, and the implementation of that theory. It is essential for anyone who is attempting to review or modify the code or who otherwise needs to understand the internal workings of the code. In addition, this document includes, in Appendix A, the System Requirements Specification for the TMAD System

  19. Origin of salt giants in abyssal serpentinite systems

    Science.gov (United States)

    Scribano, Vittorio; Carbone, Serafina; Manuella, Fabio C.; Hovland, Martin; Rueslåtten, Håkon; Johnsen, Hans-K.

    2017-10-01

    Worldwide marine salt deposits ranging over the entire geological record are generally considered climate-related evaporites, derived from the precipitation of salts (mainly chlorides and sulfates) from saturated solutions driven by solar evaporation of seawater. This explanation may be realistic for a salt thickness ≤100 m, being therefore inadequate for thicker (>1 km) deposits. Moreover, sub-seafloor salt deposits in deep marine basins are difficult to reconcile with a surface evaporation model. Marine geology reports on abyssal serpentinite systems provide an alternative explanation for some salt deposits. Seawater-driven serpentinization consumes water and increases the salinity of the associated aqueous brines. Brines can be trapped in fractures and cavities in serpentinites and the surrounding `country' rocks. Successive thermal dehydration of buried serpentinites can mobilize and accumulate the brines, forming highly saline hydrothermal solutions. These can migrate upwards and erupt onto the seafloor as saline geysers, which may form salt-saturated water pools, as are currently observed in numerous deeps in the Red Sea and elsewhere. The drainage of deep-seated saline brines to seafloor may be a long-lasting, effective process, mainly occurring in areas characterized by strong tectonic stresses and/or igneous intrusions. Alternatively, brines could be slowly expelled from fractured serpentinites by buoyancy gradients and, hence, separated salts/brines could intrude vertically into surrounding rocks, forming salt diapirs. Serpentinization is an ubiquitous, exothermic, long-lasting process which can modify large volumes of oceanic lithosphere over geological times. Therefore, buried salt deposits in many areas of the world can be reasonably related to serpentinites.

  20. SPECTRAL AMPLITUDE CODING OCDMA SYSTEMS USING ENHANCED DOUBLE WEIGHT CODE

    Directory of Open Access Journals (Sweden)

    F.N. HASOON

    2006-12-01

    Full Text Available A new code structure for spectral amplitude coding optical code division multiple access systems based on double weight (DW code families is proposed. The DW has a fixed weight of two. Enhanced double-weight (EDW code is another variation of a DW code family that can has a variable weight greater than one. The EDW code possesses ideal cross-correlation properties and exists for every natural number n. A much better performance can be provided by using the EDW code compared to the existing code such as Hadamard and Modified Frequency-Hopping (MFH codes. It has been observed that theoretical analysis and simulation for EDW is much better performance compared to Hadamard and Modified Frequency-Hopping (MFH codes.

  1. Hydrodynamic simulation of a lithium chloride salt system

    International Nuclear Information System (INIS)

    Eberle, C. S.; Herrmann, S. D.; Knighton, G. C.

    1999-01-01

    A fused lithium chloride salt system's constitutive properties were evaluated and compared to a number of fluid properties, and water was shown to be an excellent simulant of lithium chloride salt. With a simple flow model, the principal scaling term was shown to be a function of the kinematic viscosity. A water mock-up of the molten salt was also shown to be within a ±3% error in the scaling analysis. This made it possible to consider developing water scaled tests of the molten salt system. Accurate flow velocity and pressure measurements were acquired by developing a directional velocity probe. The device was constructed and calibrated with a repeatable accuracy of ±15%. This was verified by a detailed evaluation of the probe. Extensive flow measurements of the engineering scale mockup were conducted, and the results were carefully compared to radial flow patterns of a straight blade stirrer. The flow measurements demonstrated an anti-symmetric nature of the stirring, and many additional effects were also identified. The basket design was shown to prevent fluid penetration into the fuel baskets when external stirring was the flow mechanism

  2. Elements of algebraic coding systems

    CERN Document Server

    Cardoso da Rocha, Jr, Valdemar

    2014-01-01

    Elements of Algebraic Coding Systems is an introductory text to algebraic coding theory. In the first chapter, you'll gain inside knowledge of coding fundamentals, which is essential for a deeper understanding of state-of-the-art coding systems. This book is a quick reference for those who are unfamiliar with this topic, as well as for use with specific applications such as cryptography and communication. Linear error-correcting block codes through elementary principles span eleven chapters of the text. Cyclic codes, some finite field algebra, Goppa codes, algebraic decoding algorithms, and applications in public-key cryptography and secret-key cryptography are discussed, including problems and solutions at the end of each chapter. Three appendices cover the Gilbert bound and some related derivations, a derivation of the Mac- Williams' identities based on the probability of undetected error, and two important tools for algebraic decoding-namely, the finite field Fourier transform and the Euclidean algorithm f...

  3. Liquid Salt Heat Exchanger Technology for VHTR Based Applications

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Mark; Sridhara, Kumar; Allen, Todd; Peterson, Per

    2012-10-11

    The objective of this research is to evaluate performance of liquid salt fluids for use as a heat carrier for transferring high-temperature process heat from the very high-temperature reactor (VHTR) to chemical process plants. Currently, helium is being considered as the heat transfer fluid; however, the tube size requirements and the power associated with pumping helium may not be economical. Recent work on liquid salts has shown tremendous potential to transport high-temperature heat efficiently at low pressures over long distances. This project has two broad objectives: To investigate the compatibility of Incoloy 617 and coated and uncoated SiC ceramic composite with MgCl2-KCl molten salt to determine component lifetimes and aid in the design of heat exchangers and piping; and, To conduct the necessary research on the development of metallic and ceramic heat exchangers, which are needed for both the helium-to-salt side and salt-to-process side, with the goal of making these heat exchangers technologically viable. The research will consist of three separate tasks. The first task deals with material compatibility issues with liquid salt and the development of techniques for on-line measurement of corrosion products, which can be used to measure material loss in heat exchangers. Researchers will examine static corrosion of candidate materials in specific high-temperature heat transfer salt systems and develop an in situ electrochemical probe to measure metallic species concentrations dissolved in the liquid salt. The second task deals with the design of both the intermediate and process side heat exchanger systems. Researchers will optimize heat exchanger design and study issues related to corrosion, fabrication, and thermal stresses using commercial and in-house codes. The third task focuses integral testing of flowing liquid salts in a heat transfer/materials loop to determine potential issues of using the salts and to capture realistic behavior of the salts in a

  4. Comparative miRomics of Salt-Tolerant and Salt-Sensitive Rice

    Directory of Open Access Journals (Sweden)

    Goswami Kavita

    2017-06-01

    Full Text Available Increase in soil salt causes osmotic and ionic stress to plants, which inhibits their growth and productivity. Rice production is also hampered by salinity and the effect of salt is most severe at the seedling and reproductive stages. Salainity tolerance is a quantitative property controlled by multiple genes coding for signaling molecules, ion transporters, metabolic enzymes and transcription regulators. MicroRNAs are key modulators of gene-expression that act at the post-transcriptional level by translation repression or transcript cleavage. They also play an important role in regulating plant’s response to salt-stress. In this work we adopted the approach of comparative and integrated data-mining to understand the miRNA-mediated regulation of salt-stress in rice. We profiled and compared the miRNA regulations using natural varieties and transgenic lines with contrasting behaviors in response to salt-stress. The information obtained from sRNAseq, RNAseq and degradome datasets was integrated to identify the salt-deregulated miRNAs, their targets and the associated metabolic pathways. The analysis revealed the modulation of many biological pathways, which are involved in salt-tolerance and play an important role in plant phenotype and physiology. The end modifications of the miRNAs were also studied in our analysis and isomiRs having a dynamic role in salt-tolerance mechanism were identified.

  5. New rational nuclear energy system composed of accelerator molten-salt breeder (AMSB) and molten-salt power stations (MSCR)

    International Nuclear Information System (INIS)

    Furukawa, K.

    1985-01-01

    For the next century, it was predicted that some rational fission energy system breeding in significantly short doubling time less than 10 years should be developed replacing the fossil fuels. In practice, this rationality, that is, simplicity and high economy could be realized by the natural combination of: molten salt fuel concept; accelerator (spallation) breeding concept; and Thorium fuel cycle concept, in the symbiont system of Accelerator Molten-Salt breeders and Molten-Salt Power Stations. The economy of this system might significantly become better than the other breeder systems, although the prediction in Chapter 6 was too much conservative. Its more important aspect is the low cost of future R and D, which depend on the rational character of Molten-Fluoride Technology and really is verified by the basic R and D cost (only $0.13 B) in Oak Ridge N.L. It is interesting that molten-salt technology will be able to apply to chemical processing of U-Pu oxide fuels by the developing effort by USSR in near future. This fact and the demand of small power stations such as 150MWe MSCR presented here will be able to bridge between the present and the next century

  6. Candidate molten salt investigation for an accelerator driven subcritical core

    International Nuclear Information System (INIS)

    Sooby, E.; Baty, A.; Beneš, O.; McIntyre, P.; Pogue, N.; Salanne, M.; Sattarov, A.

    2013-01-01

    Highlights: • Developing accelerator driven subcritical fission to destroy transuranics in SNF. • The core is a vessel containing a molten mixture of NaCl and transuranic chlorides. • Molecular dynamics used to calculate the thermophysical properties of the salt. • Density and molecular structure for actinide salts reported here. • The neutronics of ADS fission in molten salt are presented. -- Abstract: We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated

  7. Candidate molten salt investigation for an accelerator driven subcritical core

    Energy Technology Data Exchange (ETDEWEB)

    Sooby, E., E-mail: soobyes@tamu.edu [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Baty, A. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Beneš, O. [European Commission, DG Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); McIntyre, P.; Pogue, N. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States); Salanne, M. [Université Pierre et Marie Curie, CNRS, Laboratoire PECSA, F-75005 Paris (France); Sattarov, A. [Texas A and M University, Accelerator Research Laboratory, 3380 University Dr. East, College Station, TX 77845 (United States)

    2013-09-15

    Highlights: • Developing accelerator driven subcritical fission to destroy transuranics in SNF. • The core is a vessel containing a molten mixture of NaCl and transuranic chlorides. • Molecular dynamics used to calculate the thermophysical properties of the salt. • Density and molecular structure for actinide salts reported here. • The neutronics of ADS fission in molten salt are presented. -- Abstract: We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated.

  8. Molten Salt Breeder Reactor Analysis Based on Unit Cell Model

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yongjin; Choi, Sooyoung; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    Contemporary computer codes like the MCNP6 or SCALE are only good for solving a fixed solid fuel reactor. However, due to the molten-salt fuel, MSR analysis needs some functions such as online reprocessing and refueling, and circulating fuel. J. J. Power of Oak Ridge National Laboratory (ORNL) suggested in 2013 a method for simulating the Molten Salt Breeder Reactor (MSBR) with SCALE, which does not support continuous material processing. In order to simulate MSR characteristics, the method proposes dividing a depletion time into short time intervals and batchwise reprocessing and refueling at each step. We are applying this method by using the MCNP6 and PYTHON and NEWT-TRITON-PYTHON and PYTHON code systems to MSBR. This paper contains various parameters to analyze the MSBR unit cell model such as the multiplication factor, breeding ratio, change of amount of fuel, amount of fuel feeding, and neutron flux distribution. The result of MCNP6 and NEWT module in SCALE show some difference in depletion analysis, but it still seems that they can be used to analyze MSBR. Using these two computer code system, it is possible to analyze various parameters for the MSBR unit cells such as the multiplication factor, breeding ratio, amount of material, total feeding, and neutron flux distribution. Furthermore, the two code systems will be able to be used for analyzing other MSR model or whole core models of MSR.

  9. Molten Salt Breeder Reactor Analysis Based on Unit Cell Model

    International Nuclear Information System (INIS)

    Jeong, Yongjin; Choi, Sooyoung; Lee, Deokjung

    2014-01-01

    Contemporary computer codes like the MCNP6 or SCALE are only good for solving a fixed solid fuel reactor. However, due to the molten-salt fuel, MSR analysis needs some functions such as online reprocessing and refueling, and circulating fuel. J. J. Power of Oak Ridge National Laboratory (ORNL) suggested in 2013 a method for simulating the Molten Salt Breeder Reactor (MSBR) with SCALE, which does not support continuous material processing. In order to simulate MSR characteristics, the method proposes dividing a depletion time into short time intervals and batchwise reprocessing and refueling at each step. We are applying this method by using the MCNP6 and PYTHON and NEWT-TRITON-PYTHON and PYTHON code systems to MSBR. This paper contains various parameters to analyze the MSBR unit cell model such as the multiplication factor, breeding ratio, change of amount of fuel, amount of fuel feeding, and neutron flux distribution. The result of MCNP6 and NEWT module in SCALE show some difference in depletion analysis, but it still seems that they can be used to analyze MSBR. Using these two computer code system, it is possible to analyze various parameters for the MSBR unit cells such as the multiplication factor, breeding ratio, amount of material, total feeding, and neutron flux distribution. Furthermore, the two code systems will be able to be used for analyzing other MSR model or whole core models of MSR

  10. Candidate molten salt investigation for an accelerator driven subcritical core

    Science.gov (United States)

    Sooby, E.; Baty, A.; Beneš, O.; McIntyre, P.; Pogue, N.; Salanne, M.; Sattarov, A.

    2013-09-01

    We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated. A special thanks is due to Prof. Paul Madden for introducing the ADSMS group to the concept of using the molten salt as the spallation target, rather than a conventional heavy metal spallation target. This feature helps to optimize this core as a Pu/TRU burner.

  11. Chemical interactions and thermodynamic studies in aluminum alloy/molten salt systems

    Science.gov (United States)

    Narayanan, Ramesh

    The recycling of aluminum and aluminum alloys such as Used Beverage Container (UBC) is done under a cover of molten salt flux based on (NaCl-KCl+fluorides). The reactions of aluminum alloys with molten salt fluxes have been investigated. Thermodynamic calculations are performed in the alloy/salt flux systems which allow quantitative predictions of the equilibrium compositions. There is preferential reaction of Mg in Al-Mg alloy with molten salt fluxes, especially those containing fluorides like NaF. An exchange reaction between Al-Mg alloy and molten salt flux has been demonstrated. Mg from the Al-Mg alloy transfers into the salt flux while Na from the salt flux transfers into the metal. Thermodynamic calculations indicated that the amount of Na in metal increases as the Mg content in alloy and/or NaF content in the reacting flux increases. This is an important point because small amounts of Na have a detrimental effect on the mechanical properties of the Al-Mg alloy. The reactions of Al alloys with molten salt fluxes result in the formation of bluish purple colored "streamers". It was established that the streamer is liquid alkali metal (Na and K in the case of NaCl-KCl-NaF systems) dissipating into the melt. The melts in which such streamers were observed are identified. The metal losses occurring due to reactions have been quantified, both by thermodynamic calculations and experimentally. A computer program has been developed to calculate ternary phase diagrams in molten salt systems from the constituting binary phase diagrams, based on a regular solution model. The extent of deviation of the binary systems from regular solution has been quantified. The systems investigated in which good agreement was found between the calculated and experimental phase diagrams included NaF-KF-LiF, NaCl-NaF-NaI and KNOsb3-TINOsb3-LiNOsb3. Furthermore, an insight has been provided on the interrelationship between the regular solution parameters and the topology of the phase

  12. System Based Code: Principal Concept

    International Nuclear Information System (INIS)

    Yasuhide Asada; Masanori Tashimo; Masahiro Ueta

    2002-01-01

    This paper introduces a concept of the 'System Based Code' which has initially been proposed by the authors intending to give nuclear industry a leap of progress in the system reliability, performance improvement, and cost reduction. The concept of the System Based Code intends to give a theoretical procedure to optimize the reliability of the system by administrating every related engineering requirement throughout the life of the system from design to decommissioning. (authors)

  13. An improvement study on the closed chamber distillation system for recovery of renewable salts from salt wastes containing radioactive rare earth compounds

    International Nuclear Information System (INIS)

    Eun, H.C.; Cho, Y.Z.; Lee, T.K.; Kim, I.T.; Park, G.I.; Lee, H.S.

    2013-01-01

    In this paper, an improvement study on the closed chamber distillation system for recovery of renewable salts from salt wastes containing radioactive rare earth compounds was performed to determine optimum operating conditions. It was very important to maintain the pressure in the distillation chamber below 10 Torr for a high efficiency (salt recovery >99 %) of the salt distillation. This required increasing the salt vaporization and condensation rates in the distillation system. It was confirmed that vaporization and condensation rates could be improved controlling the given temperature of top of the condensation chamber. In the distillation tests of the salt wastes containing rare earth compounds, the operation time at a given temperature was greatly reduced changing the given temperature of top of the condensation chamber from 780 to 700 deg C. (author)

  14. Expected brine movement at potential nuclear waste repository salt sites

    International Nuclear Information System (INIS)

    McCauley, V.S.; Raines, G.E.

    1987-08-01

    The BRINEMIG brine migration code predicts rates and quantities of brine migration to a waste package emplaced in a high-level nuclear waste repository in salt. The BRINEMIG code is an explicit time-marching finite-difference code that solves a mass balance equation and uses the Jenks equation to predict velocities of brine migration. Predictions were made for the seven potentially acceptable salt sites under consideration as locations for the first US high-level nuclear waste repository. Predicted total quantities of accumulated brine were on the order of 1 m 3 brine per waste package or less. Less brine accumulation is expected at domal salt sites because of the lower initial moisture contents relative to bedded salt sites. Less total accumulation of brine is predicted for spent fuel than for commercial high-level waste because of the lower temperatures generated by spent fuel. 11 refs., 36 figs., 29 tabs

  15. Laser-Induced Breakdown Spectroscopy (LIBS) in a Novel Molten Salt Aerosol System.

    Science.gov (United States)

    Williams, Ammon N; Phongikaroon, Supathorn

    2017-04-01

    In the pyrochemical separation of used nuclear fuel (UNF), fission product, rare earth, and actinide chlorides accumulate in the molten salt electrolyte over time. Measuring this salt composition in near real-time is advantageous for operational efficiency, material accountability, and nuclear safeguards. Laser-induced breakdown spectroscopy (LIBS) has been proposed and demonstrated as a potential analytical approach for molten LiCl-KCl salts. However, all the studies conducted to date have used a static surface approach which can lead to issues with splashing, low repeatability, and poor sample homogeneity. In this initial study, a novel molten salt aerosol approach has been developed and explored to measure the composition of the salt via LIBS. The functionality of the system has been demonstrated as well as a basic optimization of the laser energy and nebulizer gas pressure used. Initial results have shown that this molten salt aerosol-LIBS system has a great potential as an analytical technique for measuring the molten salt electrolyte used in this UNF reprocessing technology.

  16. Identification of a single sinusoidal bile salt uptake system in skate liver

    International Nuclear Information System (INIS)

    Fricker, G.; Hugentobler, G.; Meier, P.J.; Kurz, G.; Boyer, J.L.

    1987-01-01

    To identify the sinusoidal bile acid uptake system(s) of skate liver, photoaffinity labeling and kinetic transport studies were performed in isolated plasma membranes as well as intact hepatocytes. In both preparations photoaffinity labeling with the photolabile bile salt derivative revealed the presence of a predominant bile salt binding polypeptide with an apparent molecular weight of 54,000. The [ 3 H]-labeling of this polypeptide was inhibited by taurocholate and cholate in a concentration-dependent manner and was virtually abolished by 1 mM of the anion transport inhibitor 4,4'-diisothiocyanostilbene-2,2'-disulfonic acid. Kinetic studies of hepatic uptake with taurocholate, cholate, and the photoreactive bile salt derivative indicated the involvement of a single transport system, and all three substrates mutually competed with the uptake of each other. Finally, irreversible inhibition of the bile salt uptake system of photoaffinity labeling of hepatocytes with high concentrations of photolabile derivative reduced the V max but the K m of taurocholate uptake. These findings strongly indicate that a single polypeptide with an apparent molecular weight of 54,000 is involved in sinusoidal bile salt uptake into skate hepatocytes. These findings contrast with similar studies in rat liver that implicate both a 54,000- and 48,000-K polypeptide in bile salt uptake and are consistent with a single Na + -independent transport mechanism for hepatic bile salt uptake in this primitive vertebrate

  17. Next generation Zero-Code control system UI

    CERN Multimedia

    CERN. Geneva

    2017-01-01

    Developing ergonomic user interfaces for control systems is challenging, especially during machine upgrade and commissioning where several small changes may suddenly be required. Zero-code systems, such as *Inspector*, provide agile features for creating and maintaining control system interfaces. More so, these next generation Zero-code systems bring simplicity and uniformity and brake the boundaries between Users and Developers. In this talk we present *Inspector*, a CERN made Zero-code application development system, and we introduce the major differences and advantages of using Zero-code control systems to develop operational UI.

  18. Simulation tools and new developments of the molten salt fast reactor

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Doligez, X.; Heuer, D.; Allibert, M.; Ghetta, V.

    2010-01-01

    Starting from the Molten Salt Breeder Reactor project of Oak-Ridge, we have performed parametric studies in terms of safety coefficients, reprocessing requirements and breeding capabilities. In the frame of this major re-evaluation of the molten salt reactor (MSR), we have developed a new concept called Molten Salt Fast Reactor or MSFR, based on the Thorium fuel cycle and a fast neutron spectrum. This concept has been selected for further studies by the MSR steering committee of the Generation IV International Forum in 2009. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for materials evolution which resolves the Bateman's equations giving the population of each nucleus inside each part of the reactor at each moment. Because of MSR's fundamental characteristics compared to classical solid-fuelled reactors, the classical Bateman equations have to be modified by adding two terms representing the reprocessing capacities and the fertile or fissile alimentation. We have thus coupled neutronic and reprocessing simulation codes in a numerical tool used to calculate the extraction efficiencies of fission products, their location in the whole system (reactor and reprocessing unit) and radioprotection issues. (authors)

  19. A mean field theory of coded CDMA systems

    International Nuclear Information System (INIS)

    Yano, Toru; Tanaka, Toshiyuki; Saad, David

    2008-01-01

    We present a mean field theory of code-division multiple-access (CDMA) systems with error-control coding. On the basis of the relation between the free energy and mutual information, we obtain an analytical expression of the maximum spectral efficiency of the coded CDMA system, from which a mean-field description of the coded CDMA system is provided in terms of a bank of scalar Gaussian channels whose variances in general vary at different code symbol positions. Regular low-density parity-check (LDPC)-coded CDMA systems are also discussed as an example of the coded CDMA systems

  20. A mean field theory of coded CDMA systems

    Energy Technology Data Exchange (ETDEWEB)

    Yano, Toru [Graduate School of Science and Technology, Keio University, Hiyoshi, Kohoku-ku, Yokohama-shi, Kanagawa 223-8522 (Japan); Tanaka, Toshiyuki [Graduate School of Informatics, Kyoto University, Yoshida Hon-machi, Sakyo-ku, Kyoto-shi, Kyoto 606-8501 (Japan); Saad, David [Neural Computing Research Group, Aston University, Birmingham B4 7ET (United Kingdom)], E-mail: yano@thx.appi.keio.ac.jp

    2008-08-15

    We present a mean field theory of code-division multiple-access (CDMA) systems with error-control coding. On the basis of the relation between the free energy and mutual information, we obtain an analytical expression of the maximum spectral efficiency of the coded CDMA system, from which a mean-field description of the coded CDMA system is provided in terms of a bank of scalar Gaussian channels whose variances in general vary at different code symbol positions. Regular low-density parity-check (LDPC)-coded CDMA systems are also discussed as an example of the coded CDMA systems.

  1. SASSYS LMFBR systems analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.

    1982-01-01

    The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies

  2. An optimized symbiotic fusion and molten-salt fission reactor system

    International Nuclear Information System (INIS)

    Blinkin, V.L.; Novikov, V.M.

    A symbiotic fusion-fission reactor system which breeds nuclear fuel is discussed. In the blanket of the controlled thermonuclear reactor (CTR) uranium-233 is generated from thorium, which circulates in the form of ThF 4 mixed with molten sodium and beryllium fluorides. The molten-salt fission reactor (MSR) burns up the uranium-233 and generates tritium for the fusion reactor from lithium, which circulates in the form of LiF mixed with BeF 2 and 233 UF 4 through the MSR core. With a CTR-MSR thermal power ratio of 1:11 the system can produce electrical energy and breed fuel with a doubling time of 4-5 years. The system has the following special features: (1) Fuel reprocessing is much simpler and cheaper than for contemporary fission reactors; reprocessing consists simply in continuous removal of 233 U from the salt circulating in the CTR blanket by the fluorination method and removal of xenon from the MSR fuel salt by gas scavenging; the MSR fuel salt is periodically exchanged for fresh salt and the 233 U is then removed from it; (2) Tritium is produced in the fission reactor, which is a much simpler system than the fusion reactor; (3) The CTR blanket is almost ''clean''; no tritium is produced in it and fission fragment activity does not exceed the activity induced in the structural materials; (4) Almost all the thorium introduced into the CTR blanket can be used for producing 233 U

  3. Transient analyses for a molten salt fast reactor with optimized core geometry

    Energy Technology Data Exchange (ETDEWEB)

    Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

    2015-10-15

    Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

  4. Molten Salt Test Loop (MSTL) system customer interface document.

    Energy Technology Data Exchange (ETDEWEB)

    Gill, David Dennis; Kolb, William J.; Briggs, Ronald D.

    2013-09-01

    The National Solar Thermal Test Facility at Sandia National Laboratories has a unique test capability called the Molten Salt Test Loop (MSTL) system. MSTL is a test capability that allows customers and researchers to test components in flowing, molten nitrate salt. The components tested can range from materials samples, to individual components such as flex hoses, ball joints, and valves, up to full solar collecting systems such as central receiver panels, parabolic troughs, or linear Fresnel systems. MSTL provides realistic conditions similar to a portion of a concentrating solar power facility. The facility currently uses 60/40 nitrate %E2%80%9Csolar salt%E2%80%9D and can circulate the salt at pressure up to 40 bar (600psi), temperature to 585%C2%B0C, and flow rate of 44-50kg/s(400-600GPM) depending on temperature. The purpose of this document is to provide a basis for customers to evaluate the applicability to their testing needs, and to provide an outline of expectations for conducting testing on MSTL. The document can serve as the basis for testing agreements including Work for Others (WFO) and Cooperative Research and Development Agreements (CRADA). While this document provides the basis for these agreements and describes some of the requirements for testing using MSTL and on the site at Sandia, the document is not sufficient by itself as a test agreement. The document, however, does provide customers with a uniform set of information to begin the test planning process.

  5. Basic concept of common reactor physics code systems. Final report of working party on common reactor physics code systems (CCS)

    International Nuclear Information System (INIS)

    2004-03-01

    A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)

  6. Phosphate-dependent root system architecture responses to salt stress

    KAUST Repository

    Kawa, Dorota; Julkowska, Magdalena; Montero Sommerfeld, Hector; Horst, Anneliek ter; Haring, Michel A; Testerink, Christa

    2016-01-01

    Nutrient availability and salinity of the soil affect growth and development of plant roots. Here, we describe how phosphate availability affects root system architecture (RSA) of Arabidopsis and how phosphate levels modulate responses of the root to salt stress. Phosphate (Pi) starvation reduced main root length and increased the number of lateral roots of Arabidopsis Col-0 seedlings. In combination with salt, low Pi dampened the inhibiting effect of mild salt stress (75mM) on all measured RSA components. At higher NaCl concentrations, the Pi deprivation response prevailed over the salt stress only for lateral root elongation. The Pi deprivation response of lateral roots appeared to be oppositely affected by abscisic acid (ABA) signaling compared to the salt stress response. Natural variation in the response to the combination treatment of salt and Pi starvation within 330 Arabidopsis accessions could be grouped into four response patterns. When exposed to double stress, in general lateral roots prioritized responses to salt, while the effect on main root traits was additive. Interestingly, these patterns were not identical for all accessions studied and multiple strategies to integrate the signals from Pi deprivation and salinity were identified. By Genome Wide Association Mapping (GWAS) 13 genomic loci were identified as putative factors integrating responses to salt stress and Pi starvation. From our experiments, we conclude that Pi starvation interferes with salt responses mainly at the level of lateral roots and that large natural variation exists in the available genetic repertoire of accessions to handle the combination of stresses.

  7. Phosphate-dependent root system architecture responses to salt stress

    KAUST Repository

    Kawa, Dorota

    2016-05-20

    Nutrient availability and salinity of the soil affect growth and development of plant roots. Here, we describe how phosphate availability affects root system architecture (RSA) of Arabidopsis and how phosphate levels modulate responses of the root to salt stress. Phosphate (Pi) starvation reduced main root length and increased the number of lateral roots of Arabidopsis Col-0 seedlings. In combination with salt, low Pi dampened the inhibiting effect of mild salt stress (75mM) on all measured RSA components. At higher NaCl concentrations, the Pi deprivation response prevailed over the salt stress only for lateral root elongation. The Pi deprivation response of lateral roots appeared to be oppositely affected by abscisic acid (ABA) signaling compared to the salt stress response. Natural variation in the response to the combination treatment of salt and Pi starvation within 330 Arabidopsis accessions could be grouped into four response patterns. When exposed to double stress, in general lateral roots prioritized responses to salt, while the effect on main root traits was additive. Interestingly, these patterns were not identical for all accessions studied and multiple strategies to integrate the signals from Pi deprivation and salinity were identified. By Genome Wide Association Mapping (GWAS) 13 genomic loci were identified as putative factors integrating responses to salt stress and Pi starvation. From our experiments, we conclude that Pi starvation interferes with salt responses mainly at the level of lateral roots and that large natural variation exists in the available genetic repertoire of accessions to handle the combination of stresses.

  8. The octopus burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

    1996-09-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  9. The OCTOPUS burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

  10. The octopus burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

    1996-01-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  11. The OCTOPUS burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

  12. Transmutation and inventory analysis in an ATW molten salt system

    Energy Technology Data Exchange (ETDEWEB)

    Sisolak, J.E.; Truebenbach, M.T.; Henderson, D.L. [Univ. of Wisconsin, Madison, WI (United States)

    1995-10-01

    As an extension of earlier work to determine the equilibrium state of an ATW molten salt, power producing, reactor/transmuter, the WAIT code provides a time dependent view of material inventories and reactor parameters. By considering several cases, the authors infer that devices of this type do not reach equilibrium for dozens of years, and that equilibrium design calculations are inapplicable over most of the reactor life. Fissile inventory and k{sub eff} both vary by factors of 1.5 or more between reactor startup and ultimate convergence to equilibrium.

  13. High dynamic range coding imaging system

    Science.gov (United States)

    Wu, Renfan; Huang, Yifan; Hou, Guangqi

    2014-10-01

    We present a high dynamic range (HDR) imaging system design scheme based on coded aperture technique. This scheme can help us obtain HDR images which have extended depth of field. We adopt Sparse coding algorithm to design coded patterns. Then we utilize the sensor unit to acquire coded images under different exposure settings. With the guide of the multiple exposure parameters, a series of low dynamic range (LDR) coded images are reconstructed. We use some existing algorithms to fuse and display a HDR image by those LDR images. We build an optical simulation model and get some simulation images to verify the novel system.

  14. Concatenated coding systems employing a unit-memory convolutional code and a byte-oriented decoding algorithm

    Science.gov (United States)

    Lee, L.-N.

    1977-01-01

    Concatenated coding systems utilizing a convolutional code as the inner code and a Reed-Solomon code as the outer code are considered. In order to obtain very reliable communications over a very noisy channel with relatively modest coding complexity, it is proposed to concatenate a byte-oriented unit-memory convolutional code with an RS outer code whose symbol size is one byte. It is further proposed to utilize a real-time minimal-byte-error probability decoding algorithm, together with feedback from the outer decoder, in the decoder for the inner convolutional code. The performance of the proposed concatenated coding system is studied, and the improvement over conventional concatenated systems due to each additional feature is isolated.

  15. Status of reactor core design code system in COSINE code package

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

    2014-07-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  16. Status of reactor core design code system in COSINE code package

    International Nuclear Information System (INIS)

    Chen, Y.; Yu, H.; Liu, Z.

    2014-01-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  17. Revised SRAC code system

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Ishiguro, Yukio; Kaneko, Kunio; Ido, Masaru.

    1986-09-01

    Since the publication of JAERI-1285 in 1983 for the preliminary version of the SRAC code system, a number of additions and modifications to the functions have been made to establish an overall neutronics code system. Major points are (1) addition of JENDL-2 version of data library, (2) a direct treatment of doubly heterogeneous effect on resonance absorption, (3) a generalized Dancoff factor, (4) a cell calculation based on the fixed boundary source problem, (5) the corresponding edit required for experimental analysis and reactor design, (6) a perturbation theory calculation for reactivity change, (7) an auxiliary code for core burnup and fuel management, etc. This report is a revision of the users manual which consists of the general description, input data requirements and their explanation, detailed information on usage, mathematics, contents of libraries and sample I/O. (author)

  18. ETF system code: composition and applications

    International Nuclear Information System (INIS)

    Reid, R.L.; Wu, K.F.

    1980-01-01

    A computer code has been developed for application to ETF tokamak system and conceptual design studies. The code determines cost, performance, configuration, and technology requirements as a function of tokamak parameters. The ETF code is structured in a modular fashion in order to allow independent modeling of each major tokamak component. The primary benefit of modularization is that it allows updating of a component module, such as the TF coil module, without disturbing the remainder of the system code as long as the input/output to the modules remains unchanged. The modules may be run independently to perform specific design studies, such as determining the effect of allowable strain on TF coil structural requirements, or the modules may be executed together as a system to determine global effects, such as defining the impact of aspect ratio on the entire tokamak system

  19. Advanced thermionic reactor systems design code

    International Nuclear Information System (INIS)

    Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C.

    1991-01-01

    An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance

  20. Variable-length code construction for incoherent optical CDMA systems

    Science.gov (United States)

    Lin, Jen-Yung; Jhou, Jhih-Syue; Wen, Jyh-Horng

    2007-04-01

    The purpose of this study is to investigate the multirate transmission in fiber-optic code-division multiple-access (CDMA) networks. In this article, we propose a variable-length code construction for any existing optical orthogonal code to implement a multirate optical CDMA system (called as the multirate code system). For comparison, a multirate system where the lower-rate user sends each symbol twice is implemented and is called as the repeat code system. The repetition as an error-detection code in an ARQ scheme in the repeat code system is also investigated. Moreover, a parallel approach for the optical CDMA systems, which is proposed by Marić et al., is also compared with other systems proposed in this study. Theoretical analysis shows that the bit error probability of the proposed multirate code system is smaller than other systems, especially when the number of lower-rate users is large. Moreover, if there is at least one lower-rate user in the system, the multirate code system accommodates more users than other systems when the error probability of system is set below 10 -9.

  1. Simulation tools and new developments of the molten salt fast reactor

    International Nuclear Information System (INIS)

    Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Doligez, X.; Ghetta, V.

    2010-01-01

    In the MSFR (Molten Salt Fast Reactor), the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this design characteristic, the MSFR can thus operate with a widely varying fuel composition. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for Bateman's equations computing the population of any nucleus inside any part of the reactor at any moment. The classical Bateman's equations have been modified by adding 2 terms representing the reprocessing capacities and an online addition. We have thus coupled neutronic and reprocessing simulation codes in a numerical tool used to calculate the extraction efficiencies of fission products, their location in the whole system and radioprotection issues. The very preliminary results show the potential of the neutronic-reprocessing coupling we have developed. We also show that these studies are limited by the uncertainties on the design and the knowledge of the chemical reprocessing processes. (A.C.)

  2. Materials considerations for molten salt accelerator-based plutonium conversion systems

    International Nuclear Information System (INIS)

    DiStefano, J.R.; DeVan, J.H.; Keiser, J.R.; Klueh, R.L.; Eatherly, W.P.

    1995-03-01

    Accelerator-driven transmutation technology (ADTT) refers to a concept for a system that uses a blanket assembly driven by a source of neutrons produced when high-energy protons from an accelerator strike a heavy metal target. One application for such a system is called Accelerator-Based Plutonium Conversion, or ABC. Currently, the version of this concept being proposed by the Los Alamos National Laboratory features a liquid lead target material and a blanket fuel of molten fluorides that contain plutonium. Thus, the materials to be used in such a system must have, in addition to adequate mechanical strength, corrosion resistance to molten lead, corrosion resistance to molten fluoride salts, and resistance to radiation damage. In this report the corrosion properties of liquid lead and the LiF-BeF 2 molten salt system are reviewed in the context of candidate materials for the above application. Background information has been drawn from extensive past studies. The system operating temperature, type of protective environment, and oxidation potential of the salt are shown to be critical design considerations. Factors such as the generation of fission products and transmutation of salt components also significantly affect corrosion behavior, and procedures for inhibiting their effects are discussed. In view of the potential for extreme conditions relative to neutron fluxes and energies that can occur in an ADTT, a knowledge of radiation effects is a most important factor. Present information for potential materials selections is summarized

  3. Materials considerations for molten salt accelerator-based plutonium conversion systems

    International Nuclear Information System (INIS)

    DiStefano, J.R.; DeVan, J.H.; Keiser, J.R.; Klueh, R.L.; Eatherly, W.P.

    1995-02-01

    Accelerator-driven transmutation technology (ADTT) refers to a concept for a system that uses a blanket assembly driven by a source of neutrons produced when high-energy protons from an accelerator strike a heavy metal target. One application for such a system is called Accelerator-Based Plutonium Conversion, or ABC. Currently, the version of this concept being proposed by the Los Alamos National Laboratory features a liquid lead target material and a blanket fuel of molten fluorides that contain plutonium. Thus, the materials to be used in such a system must have, in addition to adequate mechanical strength, corrosion resistance to molten lead, corrosion resistance to molten fluoride salts, and resistance to radiation damage. In this report the corrosion properties of liquid lead and the LiF-BeF 2 molten salt system are reviewed in the context of candidate materials for the above application. Background information has been drawn from extensive past studies. The system operating temperature, type of protective environment, and oxidation potential of the salt are shown to be critical design considerations. Factors such as the generation of fission products and transmutation of salt components also significantly affect corrosion behavior, and procedures for inhibiting their effects are discussed. In view of the potential for extreme conditions relative to neutron fluxes and energies that can occur in an ADTT, a knowledge of radiation effects is a most important factor. Present information for potential materials selections is summarized

  4. Summary of computational support and general documentation for computer code (GENTREE) used in Office of Nuclear Waste Isolation Pilot Salt Site Selection Project

    International Nuclear Information System (INIS)

    Beatty, J.A.; Younker, J.L.; Rousseau, W.F.; Elayat, H.A.

    1983-01-01

    A Decision Tree Computer Model was adapted for the purposes of a Pilot Salt Site Selection Project conducted by the Office of Nuclear Waste Isolation (ONWI). A deterministic computer model was developed to structure the site selection problem with submodels reflecting the five major outcome categories (Cost, Safety, Delay, Environment, Community Impact) to be evaluated in the decision process. Time-saving modifications were made in the tree code as part of the effort. In addition, format changes allowed retention of information items which are valuable in directing future research and in isolation of key variabilities in the Site Selection Decision Model. The deterministic code was linked to the modified tree code and the entire program was transferred to the ONWI-VAX computer for future use by the ONWI project

  5. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)

  6. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).

  7. Transient freezing of molten salts in pipe-flow systems: Application to the direct reactor auxiliary cooling system (DRACS)

    International Nuclear Information System (INIS)

    Le Brun, N.; Hewitt, G.F.; Markides, C.N.

    2017-01-01

    Highlights: • A thermo-hydraulic model has been proposed to simulate the transient freezing of molten salts in complex piping systems. • The passive safety system DRACS in Generation-IV, molten salt reactor is susceptible to failure due to salt freezing. • For the prototypical 0.2 MW reactor considered in this study considerable freezing occurs after 20 minutes leading to reactor temperatures above 900 °C within 4 hours. • Conservative criteria for the most important/least known variables in the design of DRACS have been discussed. • Over-conservative approaches in designing the NDHX should be used with caution as they can promote pipe clogging due to freezing. - Abstract: The possibility of molten-salt freezing in pipe-flow systems is a key concern for the solar-energy industry and a safety issue in the new generation of molten-salt reactors, worthy of careful consideration. This paper tackles the problem of coolant solidification in complex pipe networks by developing a transient thermohydraulic model and applying it to the ‘Direct Reactor Auxiliary Cooling System’ (DRACS), the passive-safety system proposed for the Generation-IV molten-salt reactors. The results indicate that DRACS, as currently envisioned, is prone to failure due to freezing in the air/molten-salt heat exchanger, which can occur after approximately 20 minutes, leading to reactor temperatures above 900 °C within 4 hours. The occurrence of this scenario is related to an unstable behaviour mode of DRACS in which newly formed solid-salt deposit on the pipe walls acts to decrease the flow-rate in the secondary loop, facilitating additional solid-salt deposition. Conservative criteria are suggested to facilitate preliminary assessments of early-stage DRACS designs. The present study is, to the knowledge of the authors, the first of its kind in serving to illustrate possible safety concerns in molten-salt reactors, which are otherwise considered very safe in the literature. Furthermore

  8. Enhancing productivity of salt affected soils through crops and cropping system

    International Nuclear Information System (INIS)

    Singh, S.S.; Khan, A.R.

    2002-05-01

    The reclamation of salt affected soils needs the addition of soil amendment and enough water to leach down the soluble salts. The operations may also include other simple agronomic techniques to reclaim soils and to know the crops and varieties that may be grown and other management practices which may be followed on such soils (Khan, 2001). The choice of crops to be grown during reclamation of salt affected soils is very important to obtain acceptable yields. This also decides cropping systems as well as favorable diversification for early reclamation, desirable yield and to meet the other requirements of farm families. In any salt affected soils, the following three measures are adopted for reclamation and sustaining the higher productivity of reclaimed soils. 1. Suitable choice of crops, forestry and tree species; 2. Suitable choice of cropping and agroforestry system; 3. Other measures to sustain the productivity of reclaimed soils. (author)

  9. Concatenated coding system with iterated sequential inner decoding

    DEFF Research Database (Denmark)

    Jensen, Ole Riis; Paaske, Erik

    1995-01-01

    We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder......We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder...

  10. Expansion of the CHR bone code system

    International Nuclear Information System (INIS)

    Farnham, J.E.; Schlenker, R.A.

    1976-01-01

    This report describes the coding system used in the Center for Human Radiobiology (CHR) to identify individual bones and portions of bones of a complete skeletal system. It includes illustrations of various bones and bone segments with their respective code numbers. Codes are also presented for bone groups and for nonbone materials

  11. MARS code manual volume I: code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Bae, Sung Won; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Yoon, Churl

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This theory manual provides a complete list of overall information of code structure and major function of MARS including code architecture, hydrodynamic model, heat structure, trip / control system and point reactor kinetics model. Therefore, this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  12. Solidification of high temperature molten salts for thermal energy storage systems

    Science.gov (United States)

    Sheffield, J. W.

    1981-01-01

    The solidification of phase change materials for the high temperature thermal energy storage system of an advanced solar thermal power system has been examined theoretically. In light of the particular thermophysical properties of candidate phase change high temperature salts, such as the eutectic mixture of NaF - MgF2, the heat transfer characteristics of one-dimensional inward solidification for a cylindrical geometry have been studied. The Biot number for the solidified salt is shown to be the critical design parameter for constant extraction heat flux. A fin-on-fin design concept of heat transfer surface augmentation is proposed in an effort to minimize the effects of the salt's low thermal conductivity and large volume change upon fusing.

  13. System design description of forced-convection molten-salt corrosion loops MSR-FCL-3 and MSR-FCL-4

    International Nuclear Information System (INIS)

    Huntley, W.R.; Silverman, M.D.

    1976-11-01

    Molten-salt corrosion loops MSR-FCL-3 and MSR-FCL-4 are high-temperature test facilities designed to evaluate corrosion and mass transfer of modified Hastelloy N alloys for future use in Molten-Salt Breeder Reactors. Salt is circulated by a centrifugal sump pump to evaluate material compatibility with LiF-BeF 2 -ThF 4 -UF 4 fuel salt at velocities up to 6 m/s (20 fps) and at salt temperatures from 566 to 705 0 C (1050 to 1300 0 F). The report presents the design description of the various components and systems that make up each corrosion facility, such as the salt pump, corrosion specimens, salt piping, main heaters, salt coolers, salt sampling equipment, and helium cover-gas system, etc. The electrical systems and instrumentation and controls are described, and operational procedures, system limitations, and maintenance philosophy are discussed

  14. Interrelations of codes in human semiotic systems.

    OpenAIRE

    Somov, Georgij

    2016-01-01

    Codes can be viewed as mechanisms that enable relations of signs and their components, i.e., semiosis is actualized. The combinations of these relations produce new relations as new codes are building over other codes. Structures appear in the mechanisms of codes. Hence, codes can be described as transformations of structures from some material systems into others. Structures belong to different carriers, but exist in codes in their "pure" form. Building of codes over other codes fosters t...

  15. Thermal Characterization of Molten Salt Systems

    Energy Technology Data Exchange (ETDEWEB)

    Toni Y. Gutknecht; Guy L. Fredrickson

    2011-09-01

    The phase stability of molten salts in an electrorefiner (ER) may be adversely affected by the buildup of sodium, fission products, and transuranics in the electrolyte. Potential situations that need to be avoided are the following: (1) salt freezing due to an unexpected change in the liquidus temperature, (2) phase separation or non-homogeneity of the molten salt due to the precipitation of solids or formation of immiscible liquids, and (3) any mechanism that can result in the separation and concentration of fissile elements from the molten salt. Any of these situations would result in an off-normal condition outside the established safety basis for electrorefiner (ER) operations. The stability (and homogeneity) of the phases can potentially be monitored through the thermal characterization of the salts, which can be a function of impurity concentration. This report describes the experimental results of typical salts compositions, which consist of chlorides of potassium, lithium, strontium, samarium, praseodymium, lanthanum, barium, cerium, cesium, neodymium, sodium and gadolinium chlorides as a surrogate for both uranium and plutonium, used for the processing of used nuclear fuels.

  16. Automated sensitivity analysis of the radionuclide migration code UCB-NE-10.2

    International Nuclear Information System (INIS)

    Pin, F.G.; Worley, B.A.; Oblow, E.M.; Wright, R.Q.; Harper, W.V.

    1985-01-01

    The Salt Repository Project (SRP) of the U.S. Department of Energy is performing ongoing performance assessment analyses for the eventual licensing of an underground high-level nuclear waste repository in salt. As part of these studies, sensitivity and uncertainty analyses play a major role in the identification of important parameters, and in the identification of specific data needs for site characterization. Oak Ridge National Laboratory has supported the SRP in this effort resulting in thee development of an automated procedure for performing large scale sensitivity analysis using computer calculus. GRESS, GRadient Enhanced Software System, is a pre-compiler that can process FORTRAN computer codes and add derivative taking capabilities to the normal calculated results. The GRESS code is described and applied to the code UCB-NE-10.2 which simulates the migration through a sorption medium of the radionuclide members of a decay chain

  17. Savannah River Site - Salt-stone Disposal Facility Performance Assessment Update

    International Nuclear Information System (INIS)

    Newman, J.L.

    2009-01-01

    The Savannah River Site (SRS) Salt-stone Facility is currently in the midst of a Performance Assessment revision to estimate the effect on human health and the environment of adding new disposal units to the current Salt-stone Disposal Facility (SDF). These disposal units continue the ability to safely process the salt component of the radioactive liquid waste stored in the underground storage tanks at SRS, and is a crucial prerequisite for completion of the overall SRS waste disposition plan. Removal and disposal of low activity salt waste from the SRS liquid waste system is required in order to empty tanks for future tank waste processing and closure operations. The Salt-stone Production Facility (SPF) solidifies a low-activity salt stream into a grout matrix, known as salt-stone, suitable for disposal at the SDF. The ability to dispose of the low-activity salt stream in the SDF required a waste determination pursuant to Section 3116 of the Ronald Reagan National Defense Authorization Act of 2005 and was approved in January 2006. One of the requirements of Section 3116 of the NDAA is to demonstrate compliance with the performance objectives set out in Subpart C of Part 61 of Title 10, Code of Federal Regulations. The PA is the document that is used to ensure ongoing compliance. (authors)

  18. Seismic anisotropy in deforming salt bodies

    Science.gov (United States)

    Prasse, P.; Wookey, J. M.; Kendall, J. M.; Dutko, M.

    2017-12-01

    Salt is often involved in forming hydrocarbon traps. Studying salt dynamics and the deformation processes is important for the exploration industry. We have performed numerical texture simulations of single halite crystals deformed by simple shear and axial extension using the visco-plastic self consistent approach (VPSC). A methodology from subduction studies to estimate strain in a geodynamic simulation is applied to a complex high-resolution salt diapir model. The salt diapir deformation is modelled with the ELFEN software by our industrial partner Rockfield, which is based on a finite-element code. High strain areas at the bottom of the head-like strctures of the salt diapir show high amount of seismic anisotropy due to LPO development of halite crystals. The results demonstrate that a significant degree of seismic anisotropy can be generated, validating the view that this should be accounted for in the treatment of seismic data in, for example, salt diapir settings.

  19. Plotting system for the MINCS code

    International Nuclear Information System (INIS)

    Watanabe, Tadashi

    1990-08-01

    The plotting system for the MINCS code is described. The transient two-phase flow analysis code MINCS has been developed to provide a computational tool for analysing various two-phase flow phenomena in one-dimensional ducts. Two plotting systems, namely the SPLPLOT system and the SDPLOT system, can be used as the plotting functions. The SPLPLOT system is used for plotting time transients of variables, while the SDPLOT system is for spatial distributions. The SPLPLOT system is based on the SPLPACK system, which is used as a general tool for plotting results of transient analysis codes or experiments. The SDPLOT is based on the GPLP program, which is also regarded as one of the general plotting programs. In the SPLPLOT and the SDPLOT systems, the standardized data format called the SPL format is used in reading data to be plotted. The output data format of MINCS is translated into the SPL format by using the conversion system called the MINTOSPL system. In this report, how to use the plotting functions is described. (author)

  20. SRAC2006: A comprehensive neutronics calculation code system

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio; Tsuchihashi, Keichiro

    2007-02-01

    The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, S N transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)

  1. Noncoherent Spectral Optical CDMA System Using 1D Active Weight Two-Code Keying Codes

    Directory of Open Access Journals (Sweden)

    Bih-Chyun Yeh

    2016-01-01

    Full Text Available We propose a new family of one-dimensional (1D active weight two-code keying (TCK in spectral amplitude coding (SAC optical code division multiple access (OCDMA networks. We use encoding and decoding transfer functions to operate the 1D active weight TCK. The proposed structure includes an optical line terminal (OLT and optical network units (ONUs to produce the encoding and decoding codes of the proposed OLT and ONUs, respectively. The proposed ONU uses the modified cross-correlation to remove interferences from other simultaneous users, that is, the multiuser interference (MUI. When the phase-induced intensity noise (PIIN is the most important noise, the modified cross-correlation suppresses the PIIN. In the numerical results, we find that the bit error rate (BER for the proposed system using the 1D active weight TCK codes outperforms that for two other systems using the 1D M-Seq codes and 1D balanced incomplete block design (BIBD codes. The effective source power for the proposed system can achieve −10 dBm, which has less power than that for the other systems.

  2. Advanced Thermal Storage System with Novel Molten Salt: December 8, 2011 - April 30, 2013

    Energy Technology Data Exchange (ETDEWEB)

    Jonemann, M.

    2013-05-01

    Final technical progress report of Halotechnics Subcontract No. NEU-2-11979-01. Halotechnics has demonstrated an advanced thermal energy storage system with a novel molten salt operating at 700 degrees C. The molten salt and storage system will enable the use of advanced power cycles such as supercritical steam and supercritical carbon dioxide in next generation CSP plants. The salt consists of low cost, earth abundant materials.

  3. Channel coding in the space station data system network

    Science.gov (United States)

    Healy, T.

    1982-01-01

    A detailed discussion of the use of channel coding for error correction, privacy/secrecy, channel separation, and synchronization is presented. Channel coding, in one form or another, is an established and common element in data systems. No analysis and design of a major new system would fail to consider ways in which channel coding could make the system more effective. The presence of channel coding on TDRS, Shuttle, the Advanced Communication Technology Satellite Program system, the JSC-proposed Space Operations Center, and the proposed 30/20 GHz Satellite Communication System strongly support the requirement for the utilization of coding for the communications channel. The designers of the space station data system have to consider the use of channel coding.

  4. Code system BCG for gamma-ray skyshine calculation

    International Nuclear Information System (INIS)

    Ryufuku, Hiroshi; Numakunai, Takao; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

    1979-03-01

    A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)

  5. Performance optimization of spectral amplitude coding OCDMA system using new enhanced multi diagonal code

    Science.gov (United States)

    Imtiaz, Waqas A.; Ilyas, M.; Khan, Yousaf

    2016-11-01

    This paper propose a new code to optimize the performance of spectral amplitude coding-optical code division multiple access (SAC-OCDMA) system. The unique two-matrix structure of the proposed enhanced multi diagonal (EMD) code and effective correlation properties, between intended and interfering subscribers, significantly elevates the performance of SAC-OCDMA system by negating multiple access interference (MAI) and associated phase induce intensity noise (PIIN). Performance of SAC-OCDMA system based on the proposed code is thoroughly analyzed for two detection techniques through analytic and simulation analysis by referring to bit error rate (BER), signal to noise ratio (SNR) and eye patterns at the receiving end. It is shown that EMD code while using SDD technique provides high transmission capacity, reduces the receiver complexity, and provides better performance as compared to complementary subtraction detection (CSD) technique. Furthermore, analysis shows that, for a minimum acceptable BER of 10-9 , the proposed system supports 64 subscribers at data rates of up to 2 Gbps for both up-down link transmission.

  6. Fuel salt and container material studies for MOSART transforming system

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Feynberg, O.; Merzlyakov, A.; Surenkov, A.; Zagnitko, A. [National Research Center, Kurchatov Institute, Moscow (Russian Federation); Afonichkin, V.; Bovet, A.; Khokhlov, V. [Institute of High Temperature Electrochemisty, Ekaterinburg (Russian Federation); Subbotin, V.; Gordeev, M.; Panov, A.; Toropov, A. [Institute of Technical Physics, Snezhinsk (Russian Federation)

    2013-07-01

    A study is under progress to examine the feasibility of single stream Molten Salt Actinide Recycling and Transmuting system without and with Th support (MOSART) fuelled with different compositions of actinide tri-fluorides (AnF{sub 3}) from used LWR fuel. New fast-spectrum design options with homogeneous core and fuel salts with high enough solubility for AnF{sub 3} are being examined because of new goals. The flexibility of single fluid MOSART concept with Th support is underlined, particularly, possibility of its operation in self-sustainable mode (Conversion Ratio: CR=1) using different loadings and make up. The paper summarizes the most current status of fuel salt and container material data for the MOSART concept received within ISTC-3749 and ROSATOM-MARS projects. Key physical and chemical properties of various fluoride fuel salts are reported. The issues like salt purification, the electroreduction of U(IV) to U(III) in LiF-ThF{sub 4} and the electroreduction of Yb(III) to Yb(II) in LiF-NaF are detailed.

  7. Radioactive waste isolation in salt: special advisory report on the status of the Office of Nuclear Waste Isolation's plans for repository performance assessment

    International Nuclear Information System (INIS)

    Ditmars, J.D.; Walbridge, E.W.; Rote, D.M.; Harrison, W.; Herzenberg, C.L.

    1983-10-01

    Repository performance assessment is analysis that identifies events and processes that might affect a repository system for isolation of radioactive waste, examines their effects on barriers to waste migration, and estimates the probabilities of their occurrence and their consequences. In 1983 Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) prepared two plans - one for performance assessment for a waste repository in salt and one for verification and validation of performance assessment technology. At the request of the US Department of Energy's Salt Repository Project Office (SRPO), Argonne National Laboratory reviewed those plans and prepared this report to advise SRPO of specific areas where ONWI's plans for performance assessment might be improved. This report presents a framework for repository performance assessment that clearly identifies the relationships among the disposal problems, the processes underlying the problems, the tools for assessment (computer codes), and the data. In particular, the relationships among important processes and 26 model codes available to ONWI are indicated. A common suggestion for computer code verification and validation is the need for specific and unambiguous documentation of the results of performance assessment activities. A major portion of this report consists of status summaries of 27 model codes indicated as potentially useful by ONWI. The code summaries focus on three main areas: (1) the code's purpose, capabilities, and limitations; (2) status of the elements of documentation and review essential for code verification and validation; and (3) proposed application of the code for performance assessment of salt repository systems. 15 references, 6 figures, 4 tables

  8. Performance enhancement of optical code-division multiple-access systems using transposed modified Walsh code

    Science.gov (United States)

    Sikder, Somali; Ghosh, Shila

    2018-02-01

    This paper presents the construction of unipolar transposed modified Walsh code (TMWC) and analysis of its performance in optical code-division multiple-access (OCDMA) systems. Specifically, the signal-to-noise ratio, bit error rate (BER), cardinality, and spectral efficiency were investigated. The theoretical analysis demonstrated that the wavelength-hopping time-spreading system using TMWC was robust against multiple-access interference and more spectrally efficient than systems using other existing OCDMA codes. In particular, the spectral efficiency was calculated to be 1.0370 when TMWC of weight 3 was employed. The BER and eye pattern for the designed TMWC were also successfully obtained using OptiSystem simulation software. The results indicate that the proposed code design is promising for enhancing network capacity.

  9. On Analyzing LDPC Codes over Multiantenna MC-CDMA System

    Directory of Open Access Journals (Sweden)

    S. Suresh Kumar

    2014-01-01

    Full Text Available Multiantenna multicarrier code-division multiple access (MC-CDMA technique has been attracting much attention for designing future broadband wireless systems. In addition, low-density parity-check (LDPC code, a promising near-optimal error correction code, is also being widely considered in next generation communication systems. In this paper, we propose a simple method to construct a regular quasicyclic low-density parity-check (QC-LDPC code to improve the transmission performance over the precoded MC-CDMA system with limited feedback. Simulation results show that the coding gain of the proposed QC-LDPC codes is larger than that of the Reed-Solomon codes, and the performance of the multiantenna MC-CDMA system can be greatly improved by these QC-LDPC codes when the data rate is high.

  10. Coding-Spreading Tradeoff in CDMA Systems

    National Research Council Canada - National Science Library

    Bolas, Eduardo

    2002-01-01

    .... Comparing different combinations of coding and spreading with a traditional DS-CDMA, as defined in the IS-95 standard, allows the criteria to be defined for the best coding-spreading tradeoff in CDMA systems...

  11. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, Victor; Feynberg, Olga; Merzlyakov, Aleksandr; Surenkov, Aleksandr [Russian Research Center - Kurchatov Institute, Kurchatov sq. 1, Moscow, RF, 123182 (Russian Federation); Subbotin, Vladimir; Zakirov, Raul; Toropov, Andrey; Panov, Aleksandr [Russian Federal Nuclear Center - Institute of Technical Physics, Snezhinsk (Russian Federation); Afonichkin, Valery [Institute of High-Temperature Electrochemistry, Ekaterinburg (Russian Federation)

    2008-07-01

    To examine and demonstrate the feasibility of molten salt reactors (MSR) to reduce long lived waste toxicity and to produce efficiently electricity in closed fuel cycle, some national and international studies were initiated last years. In this paper main focus is placed on experimental evaluation of single stream Molten Salt Actinide Recycler and Transmuter (MOSART) system fuelled with different compositions of plutonium plus minor actinide trifluoride (AnF{sub 3}) from LWR spent nuclear fuel without U-Th support. This paper summarizes main experimental results of ISTC-1606 related to physical and chemical properties of fuel salt, container materials for fuel circuit, and fuel salt clean up of MOSART system. As result of ISTC-1606 studies claim is made, that the {sup 7}Li,Na,Be/F and {sup 7}Li,Be/F solvents selected for primary system appear to resolve main reactor physics, thermal hydraulics, materials compatibility, fuel salt clean up and safety problems as applied to the MOSART concept development. The created experimental facilities and the database on properties of fuel salt mixtures and container materials are used for a choice and improvement fuel salts and coolants for new applications of this high temperature technology for sustainable nuclear power development. (authors)

  12. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    International Nuclear Information System (INIS)

    Ignatiev, Victor; Feynberg, Olga; Merzlyakov, Aleksandr; Surenkov, Aleksandr; Subbotin, Vladimir; Zakirov, Raul; Toropov, Andrey; Panov, Aleksandr; Afonichkin, Valery

    2008-01-01

    To examine and demonstrate the feasibility of molten salt reactors (MSR) to reduce long lived waste toxicity and to produce efficiently electricity in closed fuel cycle, some national and international studies were initiated last years. In this paper main focus is placed on experimental evaluation of single stream Molten Salt Actinide Recycler and Transmuter (MOSART) system fuelled with different compositions of plutonium plus minor actinide trifluoride (AnF 3 ) from LWR spent nuclear fuel without U-Th support. This paper summarizes main experimental results of ISTC-1606 related to physical and chemical properties of fuel salt, container materials for fuel circuit, and fuel salt clean up of MOSART system. As result of ISTC-1606 studies claim is made, that the 7 Li,Na,Be/F and 7 Li,Be/F solvents selected for primary system appear to resolve main reactor physics, thermal hydraulics, materials compatibility, fuel salt clean up and safety problems as applied to the MOSART concept development. The created experimental facilities and the database on properties of fuel salt mixtures and container materials are used for a choice and improvement fuel salts and coolants for new applications of this high temperature technology for sustainable nuclear power development. (authors)

  13. Module type plant system dynamics analysis code (MSG-COPD). Code manual

    International Nuclear Information System (INIS)

    Sakai, Takaaki

    2002-11-01

    MSG-COPD is a module type plant system dynamics analysis code which involves a multi-dimensional thermal-hydraulics calculation module to analyze pool type of fast breeder reactors. Explanations of each module and the methods for the input data are described in this code manual. (author)

  14. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

    Sabek, M.; Gaafar, M.; Poucet, A.

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author)

  15. Use of computer codes for system reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Sabek, M.; Gaafar, M. (Nuclear Regulatory and Safety Centre, Atomic Energy Authority, Cairo (Egypt)); Poucet, A. (Commission of the European Communities, Ispra (Italy). Joint Research Centre)

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author).

  16. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aufiero, M.; Cammi, A.; Fiorina, C. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Leppänen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Luzzi, L., E-mail: lelio.luzzi@polimi.it [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Ricotti, M.E. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy)

    2013-10-15

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  17. Simulation tool of the on-line reprocessing unit of a molten salt reactor

    International Nuclear Information System (INIS)

    Simon, Nicole; Conocar, Olivier; Boussier, Hubert; Gastaldi, Olivier; Penit, Thomas; Walle, Eric; Huguet, Anne

    2006-01-01

    Molten salt reactor (MSR) is an interesting technology selected in the frame of the Generation IV forum. In the MSR, actinides are diluted in a molten salt which is both the fuel and the coolant. The ability of such a reactor is the reducing of the long-lived wastes due to high burn-up and the on-site simplified reprocessing directly connected with the core which preserve the salt properties necessary for its correct operation. Here is defined a flexible computer reprocessing code which can use data from neutronic calculations and can be coupled to a neutronic code. The code allow the description the whole behaviour of MSR, including, a coupled manner, both the design of the core and the optimised reprocessing scheme effects. (authors)

  18. Transient convergence and compaction of crushed salt as incorporated in the computer code EMOS

    International Nuclear Information System (INIS)

    Heijdra, J.J.; Hamilton, L.F.M.; Prij, J.; Slagter, W.

    1995-11-01

    An improved model for description of the transient convergence of cavities in rock salt, together with an improved model for the compaction of crushed salt is introduced. The covergence model is based on solutions of the analytical expressions based on secondary creep for a cylindrical and spherical cavity in rock salt. For the model for compaction of crushed salt the relations based on theoretical micro mechanisms have been fitted to laboratory results. A description is given of how the improved models are incorporated into the program EMOS. (orig.)

  19. Transient convergence and compaction of crushed salt as incorporated in the computer code EMOS

    Energy Technology Data Exchange (ETDEWEB)

    Heijdra, J.J.; Hamilton, L.F.M.; Prij, J.; Slagter, W.

    1995-11-01

    An improved model for description of the transient convergence of cavities in rock salt, together with an improved model for the compaction of crushed salt is introduced. The covergence model is based on solutions of the analytical expressions based on secondary creep for a cylindrical and spherical cavity in rock salt. For the model for compaction of crushed salt the relations based on theoretical micro mechanisms have been fitted to laboratory results. A description is given of how the improved models are incorporated into the program EMOS. (orig.).

  20. Performance assessment plans and methods for the Salt Repository Project

    International Nuclear Information System (INIS)

    1984-08-01

    This document presents the preliminary plans and anticipated methods of the Salt Repository Project (SRP) for assessing the postclosure and radiological aspects of preclosure performance of a nuclear waste repository in salt. This plan is intended to be revised on an annual basis. The emphasis in this preliminary effort is on the method of conceptually dividing the system into three subsystems (the very near field, the near field, and the far field) and applying models to analyze the behavior of each subsystem and its individual components. The next revision will contain more detailed plans being developed as part of Site Characterization Plan (SCP) activities. After a brief system description, this plan presents the performance targets which have been established for nuclear waste repositories by regulatory agencies (Chapter 3). The SRP approach to modeling, including sensitivity and uncertainty techniques is then presented (Chapter 4). This is followed by a discussion of scenario analysis (Chapter 5), a presentation of preliminary data needs as anticipated by the SRP (Chapter 6), and a presentation of the SRP approach to postclosure assessment of the very near field, the near field, and the far field (Chapters 7, 8, and 9, respectively). Preclosure radiological assessment is discussed in Chapter 10. Chapter 11 presents the SRP approach to code verification and validation. Finally, the Appendix lists all computer codes anticipated for use in performance assessments. The list of codes will be updated as plans are revised

  1. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    Energy Technology Data Exchange (ETDEWEB)

    Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

  2. Development of spent salt treatment technology by zeolite column system. Performance evaluation of zeolite column

    International Nuclear Information System (INIS)

    Miura, Hidenori; Uozumi, Koichi

    2009-01-01

    At electrorefining process, fission products(FPs) accumulate in molten salt. To avoid influence on heating control by decay heat and enlargement of FP amount in the recovered fuel, FP elements must be removed from the spent salt of the electrorefining process. For the removal of the FPs from the spent salt, we are investigating the availability of zeolite column system. For obtaining the basic data of the column system, such as flow property and ion-exchange performance while high temperature molten salt is passing through the column, and experimental apparatus equipped with fraction collector was developed. By using this apparatus, following results were obtained. 1) We cleared up the flow parameter of column system with zeolite powder, such as flow rate control by argon pressure. 2) Zeolite 4A in the column can absorb cesium that is one of the FP elements in molten salt. From these results, we got perspective on availability of the zeolite column system. (author)

  3. Design study of advanced nuclear fuel recycle system. Conceptual study of recycle system using molten salt

    International Nuclear Information System (INIS)

    Kakehi, I.; Shirai, N.; Hatano, M.; Kajitani, M.; Yonezawa, S.; Kawai, T.; Kawamura, F.; Tobe, K.; Takahashi, K.

    1996-12-01

    For the purpose of developing the future nuclear fuel recycle system, the design study of the advanced nuclear fuel recycle system is being conducted. This report describes intermediate accomplishments in the conceptual system study of the advanced nuclear fuel recycle system. Fundamental concepts of this system is the recycle system using molten salt which intend to break through the conventional concepts of purex and pellet fuel system. Contents of studies in this period are as follows, 1)feasibility study of the process by Cd-cathode for nitride fuel, 2)application study for the molten salt of low melting point (AlCl3+organic salt), 3)research for decladding (advantage of decladding by heat treatment), 4)behavior of FPs in electrorefining (behavior of iodine and volatile FP chlorides, FPs behavior in chlorination), 5)criticality analysis in electrorefiner, 6)drawing of off-gas flow diagram, 7)drawing of process machinery concept (cathode processor, vibration packing), 8)evaluation for the amounts of the high level radioactive wastes, 9)quality of the recycle fuels (FPs contamination of recycle fuel), 10)conceptual study of in-cell handling system, 11)meaning of the advanced nuclear fuel recycle system. The conceptual system study will be completed in describing concepts of the system and discussing issues for the developments. (author)

  4. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    International Nuclear Information System (INIS)

    Baratta, A.J.

    1997-01-01

    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together

  5. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Baratta, A.J.

    1997-07-01

    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

  6. System code improvements for modelling passive safety systems and their validation

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Sebastian; Cron, Daniel von der; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.

  7. Assessment of systems codes and their coupling with CFD codes in thermal–hydraulic applications to innovative reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bandini, G., E-mail: giacomino.bandini@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Polidori, M. [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Gerschenfeld, A.; Pialla, D.; Li, S. [Commissariat à l’Energie Atomique (CEA) (France); Ma, W.M.; Kudinov, P.; Jeltsov, M.; Kööp, K. [Royal Institute of Technology (KTH) (Sweden); Huber, K.; Cheng, X.; Bruzzese, C.; Class, A.G.; Prill, D.P. [Karlsruhe Institute of Technology (KIT) (Germany); Papukchiev, A. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Geffray, C.; Macian-Juan, R. [Technische Universität München (TUM) (Germany); Maas, L. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN) (France)

    2015-01-15

    Highlights: • The assessment of RELAP5, TRACE and CATHARE system codes on integral experiments is presented. • Code benchmark of CATHARE, DYN2B, and ATHLET on PHENIX natural circulation experiment. • Grid-free pool modelling based on proper orthogonal decomposition for system codes is explained. • The code coupling methodologies are explained. • The coupling of several CFD/system codes is tested against integral experiments. - Abstract: The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal–hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal–hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal–hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.

  8. Monitoring system specifications: retrieval of surf from a salt repository

    International Nuclear Information System (INIS)

    1980-01-01

    The task of developing specifications for a reference monitoring system determined by repository environmental conditions, retrieval operations, and federal regulatory criteria is discussed. The monitoring system specified in this report is capable of measuring (1) package position and orientation, (2) vault deformation, (3) brine accumulation, (4) spent fuel dissolution, (5) temperature, (6) nuclear radiation, and (7) package condition with sufficient accuracy to provide data input to a general risk assessment model. In order to define a monitoring system which can provide probabilistic data on radiological risk to operating personnel and the general public for a salt mine repository, the following information is required: (1) a complete design of the salt SURF repository including inventory, density and waste package design details; (2) probalistic failure rate data on containment integrity of the SURF waste package; (3) probabilistic failure rate data on the monitoring system components

  9. Automated sensitivity analysis of the radionuclide migration code UCBNE10.2

    International Nuclear Information System (INIS)

    Pin, F.G.; Worley, B.A.; Oblow, E.M.; Wright, R.Q.; Harper, W.V.

    1985-01-01

    The Salt Repository Project (SRP) of the US Department of Energy is performing ongoing performance assessment analyses for the eventual licensing of an underground high-level nuclear waste repository in salt. As part of these studies, sensitivity and uncertainty analysis play a major role in the identification of important parameters, and in the identification of specific data needs for site characterization. Oak Ridge National Laboratory has supported the SRP in this effort resulting in the development of an automated procedure for performing large-scale sensitivity analysis using computer calculus. GRESS, Gradient Enhanced Software System, is a pre-compiler that can process FORTRAN computer codes and add derivative taking capabilities to the normal calculated results. The GRESS code is described and applied to the code UCB-NE-10.2 which simulates the migration through an adsorptive medium of the radionuclide members of a decay chain. Conclusions are drawn relative to the applicability of GRESS for more general large-scale modeling sensitivity studies, and the role of such techniques in the overall SRP sensitivity/uncertainty program is detailed. 6 refs., 2 figs., 3 tabs

  10. Cloud-point measurement for (sulphate salts + polyethylene glycol 15000 + water) systems by the particle counting method

    International Nuclear Information System (INIS)

    Imani, A.; Modarress, H.; Eliassi, A.; Abdous, M.

    2009-01-01

    The phase separation of (water + salt + polyethylene glycol 15000) systems was studied by cloud-point measurements using the particle counting method. The effect of three kinds of sulphate salt (Na 2 SO 4 , K 2 SO 4 , (NH 4 ) 2 SO 4 ) concentration, polyethylene glycol 15000 concentration, mass ratio of polymer to salt on the cloud-point temperature of these systems have been investigated. The results obtained indicate that the cloud-point temperatures decrease linearly with increase in polyethylene glycol concentrations for different salts. Also, the cloud points decrease with an increase in mass ratio of salt to polymer.

  11. BER performance comparison of optical CDMA systems with/without turbo codes

    Science.gov (United States)

    Kulkarni, Muralidhar; Chauhan, Vijender S.; Dutta, Yashpal; Sinha, Ravindra K.

    2002-08-01

    In this paper, we have analyzed and simulated the BER performance of a turbo coded optical code-division multiple-access (TC-OCDMA) system. A performance comparison has been made between uncoded OCDMA and TC-OCDMA systems employing various OCDMA address codes (optical orthogonal codes (OOCs), Generalized Multiwavelength Prime codes (GMWPC's), and Generalized Multiwavelength Reed Solomon code (GMWRSC's)). The BER performance of TC-OCDMA systems has been analyzed and simulated by varying the code weight of address code employed by the system. From the simulation results, it is observed that lower weight address codes can be employed for TC-OCDMA systems that can have the equivalent BER performance of uncoded systems employing higher weight address codes for a fixed number of active users.

  12. Study of nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

    Ryu, Chang Mo; Kim, Yeon Seung; Eom, Heung Seop; Lee, Jong Bok; Kim, Ho Joon; Choi, Young Gil; Kim, Ko Ryeo

    1989-01-01

    Software maintenance is one of the most important problems since late 1970's.We wish to develop a nuclear computer code system to maintenance and manage KAERI's nuclear software. As a part of this system, we have developed three code management programs for use on CYBER and PC systems. They are used in systematic management of computer code in KAERI. The first program is embodied on the CYBER system to rapidly provide information on nuclear codes to the users. The second and the third programs were embodied on the PC system for the code manager and for the management of data in korean language, respectively. In the requirement analysis, we defined each code, magnetic tape, manual and abstract information data. In the conceptual design, we designed retrieval, update, and output functions. In the implementation design, we described the technical considerations of database programs, utilities, and directions for the use of databases. As a result of this research, we compiled the status of nuclear computer codes which belonged KAERI until September, 1988. Thus, by using these three database programs, we could provide the nuclear computer code information to the users more rapidly. (Author)

  13. Salt brickwork as long-term sealing in salt formations

    International Nuclear Information System (INIS)

    Walter, F.; Yaramanci, U.

    1993-01-01

    Radioactive wastes can be disposed of in deep salt formations. Rock salt is a suitable geologic medium because of its unique characteristics. Open boreholes, shafts and drifts are created to provide physical access to the repository. Long-term seals must be emplaced in these potential pathways to prevent radioactive release into the biosphere. The sealing materials must be mechanically and, most important, geochemically stable within the host rock. Salt bricks made from compressed salt-powder are understood to be the first choice long-term sealing material. Seals built of salt bricks will be ductile. Large sealing systems are built by combining the individual bricks with mortar. Raw materials for mortar are fine-grained halite powder and ground saliferous clay. This provides for the good adhesive strength of the mortar to the bricks and the high shear-strength of the mortar itself. To test the interaction of rock salt with an emplaced long-term seal, experiments will be carried out in situ, in the Asse salt mine in Germany. Simple borehole sealing experiments will be performed in horizontal holes and a complicated drift sealing experiment is planned, to demonstrate the technology of sealing a standard size drift or shaft inside a disturbed rock mass. Especially, the mechanical stability of the sealing system has to be demonstrated

  14. Modular ORIGEN-S for multi-physics code systems

    Energy Technology Data Exchange (ETDEWEB)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C., E-mail: yesilyurtg@ornl.gov, E-mail: clarnokt@ornl.gov, E-mail: gauldi@ornl.gov [Oak Ridge National Laboratory, TN (United States); Galloway, Jack, E-mail: jack@galloways.net [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2011-07-01

    The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including

  15. Modular ORIGEN-S for multi-physics code systems

    International Nuclear Information System (INIS)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C.; Galloway, Jack

    2011-01-01

    The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including

  16. Improvement of JRR-4 core management code system

    International Nuclear Information System (INIS)

    Izumo, H.; Watanabe, S.; Nagatomi, H.; Hori, N.

    2000-01-01

    In the modification of JRR-4, the fuel was changed from 93% high enrichment uranium aluminized fuel to 20% low enriched uranium silicide fuel in conformity with the framework of reduced enrichment program on JAERI research reactors. As changing of this, JRR-4 core management code system which estimates excess reactivity of core, fuel burn-up and so on, was improved too. It had been difficult for users to operate the former code system because its input-output form was text-form. But, in the new code system (COMMAS-JRR), users are able to operate the code system without using difficult text-form input. The estimation results of excess reactivity of JRR-4 LEU fuel core were showed very good agreements with the measured value. It is the strong points of this new code system to be operated simply by using the windows form pictures act on a personal workstation equip with the graphical-user-interface (GUI), and to estimate accurately the specific characteristics of the LEU core. (author)

  17. The application of LDPC code in MIMO-OFDM system

    Science.gov (United States)

    Liu, Ruian; Zeng, Beibei; Chen, Tingting; Liu, Nan; Yin, Ninghao

    2018-03-01

    The combination of MIMO and OFDM technology has become one of the key technologies of the fourth generation mobile communication., which can overcome the frequency selective fading of wireless channel, increase the system capacity and improve the frequency utilization. Error correcting coding introduced into the system can further improve its performance. LDPC (low density parity check) code is a kind of error correcting code which can improve system reliability and anti-interference ability, and the decoding is simple and easy to operate. This paper mainly discusses the application of LDPC code in MIMO-OFDM system.

  18. The PASC-3 code system and the UNIPASC environment

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Oppe, J.; Oudshoorn, H.

    1991-08-01

    A brief description is given of the PASC-3 (Petten-AMPX-SCALE) Reactor Physics code system and its associated UNIPASC work environment. The PASC-3 code system is used for criticality and reactor calculations and consists of a selection from the Oak Ridge National Laboratory AMPX-SCALE-3 code collection complemented with a number of additional codes and nuclear data bases. The original codes have been adapted to run under the UNIX operating system. The recommended nuclear data base is a complete 219 group cross section library derived from JEF-1 of which some benchmark results are presented. By the addition of the UNIPASC work environment the usage of the code system is greatly simplified, Complex chains of programs can easily be coupled together to form a single job. In addition, the model parameters can be represented by variables instead of literal values which enhances the readability and may improve the integrity of the code inputs. (author). 8 refs.; 6 figs.; 1 tab

  19. The effect of emulsifying salts on the turbidity of a diluted milk system with varying pH and protein concentration.

    Science.gov (United States)

    Culler, M D; Saricay, Y; Harte, F M

    2017-06-01

    Solutions of 10 commonly used emulsifying salts (ES) listed in the Code of Federal Regulations (21CFR133.179) for pasteurized process cheese were tested for their effect on the turbidity of a diluted milk system at different pH and protein concentrations to characterize the conditions that affect micellar structure. Emulsifying salt solutions were made by mixing the ES in a 1-in-20 dilution of water in skim milk ultrafiltrate (3 kDa molecular weight cut-off) to obtain ES concentrations from 0 to 248 mM. Skim milk was added to solutions containing nanopure water, skim milk ultrafiltrate, and a specific ES ranging in concentration from 0 to 248 mM and pH 5, 5.8, 6.8, 7.8, and 8.8. The turbidity of the samples was measured as the optical density at 400 nm immediately after mixing (time, t = 0), after 30 s (t = 30s), and after 30 min (t = 30min). Emulsifying salts were found to cause a decrease in the turbidity of the system, which was modeled using an exponential decay model, where C* represents a threshold salt concentration at which rapid dissociation occurs. At pH values 5.8 and 6.8, the ES caused the greatest decrease in turbidity of the diluted milk system. At pH 5, the ES had the least effect on the turbidity of the system. Sodium hexametaphosphate was found to have the strongest dissociative effect, with a C* value of 0.33 mM for t = 0 at pH 6.8. In contrast, the largest C* value calculated at pH 6.8 was monosodium phosphate at 278.22 mM. Increased time resulted in lower C* values. The model established for this study can be used to predict the dissociation of casein micelles in the presence of various types of ES. Copyright © 2017 American Dairy Science Association. Published by Elsevier Inc. All rights reserved.

  20. A bar coding system for environmental projects

    International Nuclear Information System (INIS)

    Barber, R.B.; Hunt, B.J.; Burgess, G.M.

    1988-01-01

    This paper presents BeCode systems, a bar coding system which provides both nuclear and commercial clients with a data capture and custody management program that is accurate, timely, and beneficial to all levels of project operations. Using bar code identifiers is an essentially paperless and error-free method which provides more efficient delivery of data through its menu card-driven structure, which speeds collection of essential data for uploading to a compatible device. The effects of this sequence include real-time information for operator analysis, management review, audits, planning, scheduling, and cost control

  1. Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1984-01-01

    Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

  2. RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1

    International Nuclear Information System (INIS)

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes

  3. Sequence Coding and Search System for licensee event reports: code listings. Volume 2

    International Nuclear Information System (INIS)

    Gallaher, R.B.; Guymon, R.H.; Mays, G.T.; Poore, W.P.; Cagle, R.J.; Harrington, K.H.; Johnson, M.P.

    1985-04-01

    Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This system provides a structured format for detailed coding of component, system, and unit effects as well as personnel errors. The database contains all current LERs submitted by nuclear power plant utilities for events occurring since 1981 and is updated on a continual basis. Volume 2 contains all valid and acceptable codes used for searching and encoding the LER data. This volume contains updated material through amendment 1 to revision 1 of the working version of ORNL/NSIC-223, Vol. 2

  4. Development of the next generation reactor analysis code system, MARBLE

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki

    2011-03-01

    A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)

  5. Study of trans-uranian incineration in molten salt reactor; Etude de l'incineration des transuraniens en reacteur a sel fondu

    Energy Technology Data Exchange (ETDEWEB)

    Valade, M

    2000-10-27

    With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

  6. Development of realistic thermal hydraulic system analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

  7. Development of realistic thermal hydraulic system analysis code

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, B. D; Kim, K. D.

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

  8. Integrated Validation System for a Thermal-hydraulic System Code, TASS/SMR-S

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee-Kyung; Kim, Hyungjun; Kim, Soo Hyoung; Hwang, Young-Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Hyeon-Soo [Chungnam National University, Daejeon (Korea, Republic of)

    2015-10-15

    Development including enhancement and modification of thermal-hydraulic system computer code is indispensable to a new reactor, SMART. Usually, a thermal-hydraulic system code validation is achieved by a comparison with the results of corresponding physical effect tests. In the reactor safety field, a similar concept, referred to as separate effect tests has been used for a long time. But there are so many test data for comparison because a lot of separate effect tests and integral effect tests are required for a code validation. It is not easy to a code developer to validate a computer code whenever a code modification is occurred. IVS produces graphs which shown the comparison the code calculation results with the corresponding test results automatically. IVS was developed for a validation of TASS/SMR-S code. The code validation could be achieved by a comparison code calculation results with corresponding test results. This comparison was represented as a graph for convenience. IVS is useful before release a new code version. The code developer can validate code result easily using IVS. Even during code development, IVS could be used for validation of code modification. The code developer could gain a confidence about his code modification easily and fast and could be free from tedious and long validation work. The popular software introduced in IVS supplies better usability and portability.

  9. Customer interface document for the Molten Salt Test Loop (MSTL) system.

    Energy Technology Data Exchange (ETDEWEB)

    Pettit, Kathleen; Kolb, William J.; Gill, David Dennis; Briggs, Ronald D.

    2012-03-01

    The National Solar Thermal Test Facility at Sandia National Laboratories has a unique test capability called the Molten Salt Test Loop (MSTL) system. MSTL is a test capability that allows customers and researchers to test components in flowing, molten nitrate salt. The components tested can range from materials samples, to individual components such as flex hoses, ball joints, and valves, up to full solar collecting systems such as central receiver panels, parabolic troughs, or linear Fresnel systems. MSTL provides realistic conditions similar to a portion of a concentrating solar power facility. The facility currently uses 60/40 nitrate 'solar salt' and can circulate the salt at pressure up to 600psi, temperature to 585 C, and flow rate of 400-600GPM depending on temperature. The purpose of this document is to provide a basis for customers to evaluate the applicability to their testing needs, and to provide an outline of expectations for conducting testing on MSTL. The document can serve as the basis for testing agreements including Work for Others (WFO) and Cooperative Research and Development Agreements (CRADA). While this document provides the basis for these agreements and describes some of the requirements for testing using MSTL and on the site at Sandia, the document is not sufficient by itself as a test agreement. The document, however, does provide customers with a uniform set of information to begin the test planning process.

  10. Bar-code automated waste tracking system

    International Nuclear Information System (INIS)

    Hull, T.E.

    1994-10-01

    The Bar-Code Automated Waste Tracking System was designed to be a site-Specific program with a general purpose application for transportability to other facilities. The system is user-friendly, totally automated, and incorporates the use of a drive-up window that is close to the areas dealing in container preparation, delivery, pickup, and disposal. The system features ''stop-and-go'' operation rather than a long, tedious, error-prone manual entry. The system is designed for automation but allows operators to concentrate on proper handling of waste while maintaining manual entry of data as a backup. A large wall plaque filled with bar-code labels is used to input specific details about any movement of waste

  11. Joint design of QC-LDPC codes for coded cooperation system with joint iterative decoding

    Science.gov (United States)

    Zhang, Shunwai; Yang, Fengfan; Tang, Lei; Ejaz, Saqib; Luo, Lin; Maharaj, B. T.

    2016-03-01

    In this paper, we investigate joint design of quasi-cyclic low-density-parity-check (QC-LDPC) codes for coded cooperation system with joint iterative decoding in the destination. First, QC-LDPC codes based on the base matrix and exponent matrix are introduced, and then we describe two types of girth-4 cycles in QC-LDPC codes employed by the source and relay. In the equivalent parity-check matrix corresponding to the jointly designed QC-LDPC codes employed by the source and relay, all girth-4 cycles including both type I and type II are cancelled. Theoretical analysis and numerical simulations show that the jointly designed QC-LDPC coded cooperation well combines cooperation gain and channel coding gain, and outperforms the coded non-cooperation under the same conditions. Furthermore, the bit error rate performance of the coded cooperation employing jointly designed QC-LDPC codes is better than those of random LDPC codes and separately designed QC-LDPC codes over AWGN channels.

  12. Source Code Vulnerabilities in IoT Software Systems

    Directory of Open Access Journals (Sweden)

    Saleh Mohamed Alnaeli

    2017-08-01

    Full Text Available An empirical study that examines the usage of known vulnerable statements in software systems developed in C/C++ and used for IoT is presented. The study is conducted on 18 open source systems comprised of millions of lines of code and containing thousands of files. Static analysis methods are applied to each system to determine the number of unsafe commands (e.g., strcpy, strcmp, and strlen that are well-known among research communities to cause potential risks and security concerns, thereby decreasing a system’s robustness and quality. These unsafe statements are banned by many companies (e.g., Microsoft. The use of these commands should be avoided from the start when writing code and should be removed from legacy code over time as recommended by new C/C++ language standards. Each system is analyzed and the distribution of the known unsafe commands is presented. Historical trends in the usage of the unsafe commands of 7 of the systems are presented to show how the studied systems evolved over time with respect to the vulnerable code. The results show that the most prevalent unsafe command used for most systems is memcpy, followed by strlen. These results can be used to help train software developers on secure coding practices so that they can write higher quality software systems.

  13. Thorium cycle and molten salt reactors: field parameters and field constraints investigations toward 'thorium molten salt reactor' definition

    International Nuclear Information System (INIS)

    Mathieu, L.

    2005-09-01

    Producing nuclear energy in order to reduce the anthropic CO 2 emission requires major technological advances. Nuclear plants of 4. generation have to respond to several constraints, as safety improvements, fuel breeding and radioactive waste minimization. For this purpose, it seems promising to use Thorium Cycle in Molten Salt Reactors. Studies on this domain have already been carried out. However, the final concept suffered from serious issues and was discontinued. A new reflection on this topic is being led in order to find acceptable solutions, and to design the Thorium Molten Salt Reactor concept. A nuclear reactor is simulated by the coupling of a neutron transport code with a materials evolution code. This allows us to reproduce the reactor behavior and its evolution all along its operation. Thanks to this method, we have studied a large number of reactor configurations. We have evaluated their efficiency through a group of constraints they have to satisfy. This work leads us to a better understanding of many physical phenomena controlling the reactor behavior. As a consequence, several efficient configurations have been discovered, allowing the emergence of new points of view in the research of Molten Salt Reactors. (author)

  14. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

    Sabek, M.; Gaafar, M.; Poucet, A.

    1988-01-01

    This paper gives a collective summary of the studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRANTIC, FTAP, computer code package RALLY, and BOUNDS codes. Two reference study cases were executed by each code. The results obtained logic/probabilistic analysis as well as computation time are compared

  15. Development of fuel cycle technology for molten-salt reactor systems

    International Nuclear Information System (INIS)

    Uhlir, J.

    2006-01-01

    Full text: Full text: The Molten-Salt Reactor (MSR) represents one of promising advanced reactor type assigned to the GEN IV reactor systems. It can be operated either as thorium breeder within the Th -133U fuel cycle or as actinide transmuter incinerating transuranium fuel. Essentially the main advantage of MSR comes out from the prerequisite, that this reactor type should be directly connected with the 'on-line' reprocessing of circulating liquid (molten-salt) fuel. This principle should allow very effective extraction of freshly constituted fissile material (233U). Besides, the on-line fuel salt clean up is necessary within a long run to keep the reactor in operation. As a matter of principle, it permits to clear away typical reactor poisons like xenon, krypton, lanthanides etc. and possibly also other products of burned plutonium and transmuted minor actinides. The fuel salt clean up technology should be linked with the fresh MSR fuel processing to continuously refill the new fuel (thorium or transuranics) into the reactor system. On the other hand, the technologies of fresh transuranium molten-salt fuel processing from the current LWR spent fuel and of the on-line reprocessing of MSR fuel represent two killing points of the whole MSR technology, which have to be successfully solved before MSR deployment in the future. There are three main pyrochemical partitioning techniques proposed for processing and/or reprocessing of MSR fuel: Fluoride volatilization processes, Molten salt / liquid metal extraction processes and Electrochemical separation processes. Two of them - Fluoride Volatility Method and Electrochemical separation process from fluoride media are under development in the Nuclear Research Institute Rez pic. R and D in the field of Fluoride Volatility Method is concentrated to the development and verification of experimental semi-pilot technology for LWR spent fuel reprocessing, which may result in a product the form and composition of which might be

  16. Plutonium and americium recovery from spent molten-salt-extraction salts with aluminum-magnesium alloys

    International Nuclear Information System (INIS)

    Cusick, M.J.; Sherwood, W.G.; Fitzpatrick, R.F.

    1984-01-01

    Development work was performed to determine the feasibility of removing plutonium and americium from spent molten-salt-extraction (MSE) salts using Al-Mg alloys. If the product buttons from this process are compatible with subsequent aqueous processing, the complex chloride-to-nitrate aqueous conversion step which is presently required for these salts may be eliminated. The optimum alloy composition used to treat spent 8 wt % MSE salts in the past yielded poor phase-disengagement characteristics when applied to 30 mol % salts. After a limited investigation of other alloy compositions in the Al-Mg-Pu-Am system, it was determined that the Al-Pu-Am system could yield a compatible alloy. In this system, experiments were performed to investigate the effects of plutonium loading in the alloy, excess magnesium, age of the spent salt on actinide recovery, phase disengagement, and button homogeneity. Experimental results indicate that 95 percent plutonium recoveries can be attained for fresh salts. Further development is required for backlog salts generated prior to 1981. A homogeneous product alloy, as required for aqueous processing, could not be produced

  17. Web- and system-code based, interactive, nuclear power plant simulators

    International Nuclear Information System (INIS)

    Kim, K. D.; Jain, P.; Rizwan, U.

    2006-01-01

    Using two different approaches, on-line, web- and system-code based graphical user interfaces have been developed for reactor system analysis. Both are LabVIEW (graphical programming language developed by National Instruments) based systems that allow local users as well as those at remote sites to run, interact and view the results of the system code in a web browser. In the first approach, only the data written by the system code in a tab separated ASCII output file is accessed and displayed graphically. In the second approach, LabVIEW virtual instruments are coupled with the system code as dynamic link libraries (DLL). RELAP5 is used as the system code to demonstrate the capabilities of these approaches. From collaborative projects between teams in geographically remote locations to providing system code experience to distance education students, these tools can be very beneficial in many areas of teaching and R and D. (authors)

  18. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost

  19. Development of an advanced code system for fast-reactor transient analysis

    International Nuclear Information System (INIS)

    Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

    2005-01-01

    FAST (Fast-spectrum Advanced Systems for power production and resource management) is a recently approved PSI activity in the area of fast spectrum core and safety analysis with emphasis on generic developments and Generation IV systems. In frames of the FAST project we will study both statics and transients core physics, reactor system behaviour and safety; related international experiments. The main current goal of the project is to develop unique analytical and code capability for core and safety analysis of critical (and sub-critical) fast spectrum systems with an initial emphasis on a gas cooled fast reactors. A structure of the code system is shown on Fig. 1. The main components of the FAST code system are 1) ERANOS code for preparation of basic x-sections and their partial derivatives; 2) PARCS transient nodal-method multi-group neutron diffusion code for simulation of spatial (3D) neutron kinetics in hexagonal and square geometries; 3) TRAC/AAA code for system thermal hydraulics; 4) FRED transient model for fuel thermal-mechanical behaviour; 5) PVM system as an interface between separate parts of the code system. The paper presents a structure of the code system (Fig. 1), organization of interfaces and data exchanges between main parts of the code system, examples of verification and application of separate codes and the system as a whole. (authors)

  20. Structure and thermodynamics of molten salts

    International Nuclear Information System (INIS)

    Papatheodorou, G.N.

    1983-01-01

    This chapter investigates single-component molten salts and multicomponent salt mixtures. Molten salts provide an important testing ground for theories of liquids, solutions, and plasmas. Topics considered include molten salts as liquids (the pair potential, the radial distribution function, methods of characterization), single salts (structure, thermodynamic correlations), and salt mixtures (the thermodynamics of mixing; spectroscopy and structure). Neutron and X-ray scattering techniques are used to determine the structure of molten metal halide salts. The corresponding-states theory is used to obtain thermodynamic correlations on single salts. Structural information on salt mixtures is obtained by using vibrational (Raman) and electronic absorption spectroscopy. Charge-symmetrical systems and charge-unsymmetrical systems are used to examine the thermodynamics of salt mixtures

  1. HTGR molten salt sensible energy transmission and storage system design and costs

    International Nuclear Information System (INIS)

    1981-09-01

    This report, which was prepared for Gas-Cooled Reactor Associates by United Engineers and Constructors under Contract No. GCRA/UE and C 81-203, presents the design and cost for a molten salt Sensible Energy Transmission and Storage (SETS) System. Although the reference system for this study is sized to be compatible with an 1170 MW(t) HTGR Nuclear Heat Source, the results and conclusions should be generally applicable to most large scale molten salt energy transmission system applications. A preliminary conceptual design is presented and alternative configurations are discussed. The sensitivity of system costs to variations in important system parameters are also presented. Costs for a reference case conceptual design are reported in constant 1980 dollars and inflated 1995 dollars

  2. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.

    1983-04-01

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  3. Performance enhancement of successive interference cancellation scheme based on spectral amplitude coding for optical code-division multiple-access systems using Hadamard codes

    Science.gov (United States)

    Eltaif, Tawfig; Shalaby, Hossam M. H.; Shaari, Sahbudin; Hamarsheh, Mohammad M. N.

    2009-04-01

    A successive interference cancellation scheme is applied to optical code-division multiple-access (OCDMA) systems with spectral amplitude coding (SAC). A detailed analysis of this system, with Hadamard codes used as signature sequences, is presented. The system can easily remove the effect of the strongest signal at each stage of the cancellation process. In addition, simulation of the prose system is performed in order to validate the theoretical results. The system shows a small bit error rate at a large number of active users compared to the SAC OCDMA system. Our results reveal that the proposed system is efficient in eliminating the effect of the multiple-user interference and in the enhancement of the overall performance.

  4. Study of trans-uranian incineration in molten salt reactor; Etude de l'incineration des transuraniens en reacteur a sel fondu

    Energy Technology Data Exchange (ETDEWEB)

    Valade, M

    2000-10-27

    With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

  5. ATHENA code manual. Volume 1. Code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Carlson, K.E.; Roth, P.A.; Ransom, V.H.

    1986-09-01

    The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems which may be found in fusion reactors, space reactors, and other advanced systems. A generic modeling approach is utilized which permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of a complete facility. Several working fluids are available to be used in one or more interacting loops. Different loops may have different fluids with thermal connections between loops. The modeling theory and associated numerical schemes are documented in Volume I in order to acquaint the user with the modeling base and thus aid effective use of the code. The second volume contains detailed instructions for input data preparation

  6. Development of new two-dimensional spectral/spatial code based on dynamic cyclic shift code for OCDMA system

    Science.gov (United States)

    Jellali, Nabiha; Najjar, Monia; Ferchichi, Moez; Rezig, Houria

    2017-07-01

    In this paper, a new two-dimensional spectral/spatial codes family, named two dimensional dynamic cyclic shift codes (2D-DCS) is introduced. The 2D-DCS codes are derived from the dynamic cyclic shift code for the spectral and spatial coding. The proposed system can fully eliminate the multiple access interference (MAI) by using the MAI cancellation property. The effect of shot noise, phase-induced intensity noise and thermal noise are used to analyze the code performance. In comparison with existing two dimensional (2D) codes, such as 2D perfect difference (2D-PD), 2D Extended Enhanced Double Weight (2D-Extended-EDW) and 2D hybrid (2D-FCC/MDW) codes, the numerical results show that our proposed codes have the best performance. By keeping the same code length and increasing the spatial code, the performance of our 2D-DCS system is enhanced: it provides higher data rates while using lower transmitted power and a smaller spectral width.

  7. HELIAS module development for systems codes

    Energy Technology Data Exchange (ETDEWEB)

    Warmer, F., E-mail: Felix.Warmer@ipp.mpg.de; Beidler, C.D.; Dinklage, A.; Egorov, K.; Feng, Y.; Geiger, J.; Schauer, F.; Turkin, Y.; Wolf, R.; Xanthopoulos, P.

    2015-02-15

    In order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations. This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work. This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs.

  8. NALAP: an LMFBR system transient code

    International Nuclear Information System (INIS)

    Martin, B.A.; Agrawal, A.K.; Albright, D.C.; Epel, L.G.; Maise, G.

    1975-07-01

    NALAP is a LMFBR system transient code. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic response of sodium cooled fast breeder reactors when subjected to postulated accidents such as a massive pipe break as well as a variety of other upset conditions that do not disrupt the system geometry. Various components of the plant are represented by control volumes. These control volumes are connected by junctions some of which may be leak or fill junctions. The fluid flow equations are modeled as compressible, single-stream flow with momentum flux in one dimension. The transient response is computed by integrating the thermal-hydraulic conservation equations from user-initialized operating conditions by an implicit numerical scheme. Point kinetics approximation is used to represent the time dependent heat generation in the reactor core

  9. Channel coding for underwater acoustic single-carrier CDMA communication system

    Science.gov (United States)

    Liu, Lanjun; Zhang, Yonglei; Zhang, Pengcheng; Zhou, Lin; Niu, Jiong

    2017-01-01

    CDMA is an effective multiple access protocol for underwater acoustic networks, and channel coding can effectively reduce the bit error rate (BER) of the underwater acoustic communication system. For the requirements of underwater acoustic mobile networks based on CDMA, an underwater acoustic single-carrier CDMA communication system (UWA/SCCDMA) based on the direct-sequence spread spectrum is proposed, and its channel coding scheme is studied based on convolution, RA, Turbo and LDPC coding respectively. The implementation steps of the Viterbi algorithm of convolutional coding, BP and minimum sum algorithms of RA coding, Log-MAP and SOVA algorithms of Turbo coding, and sum-product algorithm of LDPC coding are given. An UWA/SCCDMA simulation system based on Matlab is designed. Simulation results show that the UWA/SCCDMA based on RA, Turbo and LDPC coding have good performance such that the communication BER is all less than 10-6 in the underwater acoustic channel with low signal to noise ratio (SNR) from -12 dB to -10dB, which is about 2 orders of magnitude lower than that of the convolutional coding. The system based on Turbo coding with Log-MAP algorithm has the best performance.

  10. Analysis of a molten salt reactor benchmark

    International Nuclear Information System (INIS)

    Ghosh, Biplab; Bajpai, Anil; Degweker, S.B.

    2013-01-01

    This paper discusses results of our studies of an IAEA molten salt reactor (MSR) benchmark. The benchmark, proposed by Japan, involves burnup calculations of a single lattice cell of a MSR for burning plutonium and other minor actinides. We have analyzed this cell with in-house developed burnup codes BURNTRAN and McBURN. This paper also presents a comparison of the results of our codes and those obtained by the proposers of the benchmark. (author)

  11. A Robust Cross Coding Scheme for OFDM Systems

    NARCIS (Netherlands)

    Shao, X.; Slump, Cornelis H.

    2010-01-01

    In wireless OFDM-based systems, coding jointly over all the sub-carriers simultaneously performs better than coding separately per sub-carrier. However, the joint coding is not always optimal because its achievable channel capacity (i.e. the maximum data rate) is inversely proportional to the

  12. 42 CFR 405.512 - Carriers' procedural terminology and coding systems.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 2 2010-10-01 2010-10-01 false Carriers' procedural terminology and coding systems... Determining Reasonable Charges § 405.512 Carriers' procedural terminology and coding systems. (a) General. Procedural terminology and coding systems are designed to provide physicians and third party payers with a...

  13. Generalized optical code construction for enhanced and Modified Double Weight like codes without mapping for SAC-OCDMA systems

    Science.gov (United States)

    Kumawat, Soma; Ravi Kumar, M.

    2016-07-01

    Double Weight (DW) code family is one of the coding schemes proposed for Spectral Amplitude Coding-Optical Code Division Multiple Access (SAC-OCDMA) systems. Modified Double Weight (MDW) code for even weights and Enhanced Double Weight (EDW) code for odd weights are two algorithms extending the use of DW code for SAC-OCDMA systems. The above mentioned codes use mapping technique to provide codes for higher number of users. A new generalized algorithm to construct EDW and MDW like codes without mapping for any weight greater than 2 is proposed. A single code construction algorithm gives same length increment, Bit Error Rate (BER) calculation and other properties for all weights greater than 2. Algorithm first constructs a generalized basic matrix which is repeated in a different way to produce the codes for all users (different from mapping). The generalized code is analysed for BER using balanced detection and direct detection techniques.

  14. Simplified modeling and code usage in the PASC-3 code system by the introduction of a programming environment

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Oppe, J.; Oudshoorn, H.L.; Slobben, J.

    1991-06-01

    A brief description is given of the PASC-3 (Petten-AMPX-SCALE) Reactor Physics code system and associated UNIPASC work environment. The PASC-3 code system is used for criticality and reactor calculations and consists of a selection from the Oak Ridge National Laboratory AMPX-SCALE-3 code collection complemented with a number of additional codes and nuclear data bases. The original codes have been adapted to run under the UNIX operating system. The recommended nuclear data base is a complete 219 group cross section library derived from JEF-1 of which some benchmark results are presented. By the addition of the UNIPASC work environment the usage of the code system is greatly simplified. Complex chains of programs can easily be coupled together to form a single job. In addition, the model parameters can be represented by variables instead of literal values which enhances the readability and may improve the integrity of the code inputs. (author). 8 refs.; 6 figs.; 1 tab

  15. Arabic Natural Language Processing System Code Library

    Science.gov (United States)

    2014-06-01

    Adelphi, MD 20783-1197 This technical note provides a brief description of a Java library for Arabic natural language processing ( NLP ) containing code...for training and applying the Arabic NLP system described in the paper "A Cross-Task Flexible Transition Model for Arabic Tokenization, Affix...and also English) natural language processing ( NLP ), containing code for training and applying the Arabic NLP system described in Stephen Tratz’s

  16. Fast decoding algorithms for coded aperture systems

    International Nuclear Information System (INIS)

    Byard, Kevin

    2014-01-01

    Fast decoding algorithms are described for a number of established coded aperture systems. The fast decoding algorithms for all these systems offer significant reductions in the number of calculations required when reconstructing images formed by a coded aperture system and hence require less computation time to produce the images. The algorithms may therefore be of use in applications that require fast image reconstruction, such as near real-time nuclear medicine and location of hazardous radioactive spillage. Experimental tests confirm the efficacy of the fast decoding techniques

  17. Collection and evaluation of salt mixing data with the real time data acquisition system. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Glazer, S.; Chiu, C.; Todreas, N.E.

    1977-09-01

    A minicomputer based real time data acquisition system was designed and built to facilitate data collection during salt mixing tests in mock ups of LMFBR rod bundles. The system represents an expansion of data collection capabilities over previous equipment. It performs steady state and transient monitoring and recording of up to 512 individual electrical resistance probes. Extensive real time software was written to govern all phases of the data collection procedure, including probe definition, probe calibration, salt mixing test data acquisition and storage, and data editing. Offline software was also written to permit data examination and reduction to dimensionless salt concentration maps. Finally, the computer program SUPERENERGY was modified to permit rapid extraction of parameters from dimensionless salt concentration maps. The document describes the computer system, and includes circuit diagrams of all custom built components. It also includes descriptions and listings of all software written, as well as extensive user instructions.

  18. Temperature distributions in a salt formation used for the ultimate disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Ploumen, P.

    1980-01-01

    In the Federal Republic of Germany the works on waste disposal is focussed on the utilization of a salt formation for ultimate disposal of radioactive wastes. Heat released from the high-level waste will be dissipated in the salt and the surrounding geologic formations. The occuring temperature distributions will be calculated with computer codes. A survey of the developed computer codes will be shown; the results for a selected example, taking into account the loading sequence of the waste, the mine ventilation as well as an air gap between the waste and the salt, will be discussed. Furthermore it will be shown that by varying the disposal parameters, the maximum salt temperature can be below any described value. (Auth.)

  19. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

    1982-12-01

    Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

  20. Molten Salt-Carbon Nanotube Thermal Energy Storage for Concentrating Solar Power Systems Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Michael Schuller; Frank Little; Darren Malik; Matt Betts; Qian Shao; Jun Luo; Wan Zhong; Sandhya Shankar; Ashwin Padmanaban

    2012-03-30

    We demonstrated that adding nanoparticles to a molten salt would increase its utility as a thermal energy storage medium for a concentrating solar power system. Specifically, we demonstrated that we could increase the specific heat of nitrate and carbonate salts containing 1% or less of alumina nanoparticles. We fabricated the composite materials using both evaporative and air drying methods. We tested several thermophysical properties of the composite materials, including the specific heat, thermal conductivity, latent heat, and melting point. We also assessed the stability of the composite material with repeated thermal cycling and the effects of adding the nanoparticles on the corrosion of stainless steel by the composite salt. Our results indicate that stable, repeatable 25-50% improvements in specific heat are possible for these materials. We found that using these composite salts as the thermal energy storage material for a concentrating solar thermal power system can reduce the levelized cost of electricity by 10-20%. We conclude that these materials are worth further development and inclusion in future concentrating solar power systems.

  1. Locally Minimum Storage Regenerating Codes in Distributed Cloud Storage Systems

    Institute of Scientific and Technical Information of China (English)

    Jing Wang; Wei Luo; Wei Liang; Xiangyang Liu; Xiaodai Dong

    2017-01-01

    In distributed cloud storage sys-tems, inevitably there exist multiple node fail-ures at the same time. The existing methods of regenerating codes, including minimum storage regenerating (MSR) codes and mini-mum bandwidth regenerating (MBR) codes, are mainly to repair one single or several failed nodes, unable to meet the repair need of distributed cloud storage systems. In this paper, we present locally minimum storage re-generating (LMSR) codes to recover multiple failed nodes at the same time. Specifically, the nodes in distributed cloud storage systems are divided into multiple local groups, and in each local group (4, 2) or (5, 3) MSR codes are constructed. Moreover, the grouping method of storage nodes and the repairing process of failed nodes in local groups are studied. The-oretical analysis shows that LMSR codes can achieve the same storage overhead as MSR codes. Furthermore, we verify by means of simulation that, compared with MSR codes, LMSR codes can reduce the repair bandwidth and disk I/O overhead effectively.

  2. SWAT3.1 - the integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP

    International Nuclear Information System (INIS)

    Suyama, Kenya; Mochizuki, Hiroki; Takada, Tomoyuki; Ryufuku, Susumu; Okuno, Hiroshi; Murazaki, Minoru; Ohkubo, Kiyoshi

    2009-05-01

    Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC,which is widely used in Japan, and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinides and the fission products in the spent nuclear fuel. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP, and ORIGEN2. This enables us to treat the arbitrary fuel geometry and to generate the effective cross section data to be used in the burnup calculation with few approximations. This report describes the outline, input data instruction and several examples of the calculation. (author)

  3. Hydrological methods preferentially recover cesium from nuclear waste salt cake

    International Nuclear Information System (INIS)

    Brooke, J.N.; Hamm, L.L.

    1997-01-01

    The Savannah River Site is treating high level radioactive waste in the form of insoluble solids (sludge), crystallized salt (salt cake), and salt solutions. High costs and operational concerns have prompted DOE to look for ways to improve the salt cake treatment process. A numerical model was developed to evaluate the feasibility of pump and treat technology for extracting cesium from salt cake. A modified version of the VAM3DCG code was used to first establish a steady-state flow field, then to simulate 30 days of operation. Simulation results suggest that efficient cesium extraction can be obtained with low displacement volumes. The actual extraction process will probably be less impressive because of nonuniform properties. 2 refs., 2 figs

  4. JEMs and incompatible occupational coding systems: Effect of manual and automatic recoding of job codes on exposure assignment

    NARCIS (Netherlands)

    Koeman, T.; Offermans, N.S.M.; Christopher-De Vries, Y.; Slottje, P.; Brandt, P.A. van den; Goldbohm, R.A.; Kromhout, H.; Vermeulen, R.

    2013-01-01

    Background: In epidemiological studies, occupational exposure estimates are often assigned through linkage of job histories to job-exposure matrices (JEMs). However, available JEMs may have a coding system incompatible with the coding system used to code the job histories, necessitating a

  5. Safety evaluation of geological disposal concepts for low and medium-level wastes in rock-salt (Pacoma project)

    International Nuclear Information System (INIS)

    Prij, J.; Van Dalen, A.; Roodbergen, H.A.; Slagter, W.; Van Weers, A.W.; Zanstra, D.A.; Glasbergen, P.; Koester, H.W.; Lembrechts, J.F.; Nijhof-Pan, I.; Slot, A.F.M.

    1991-01-01

    In the framework of the Performance Assessment of Confinements for MLW and Alpha Waste (PACOMA) the disposal options dealing with rock-salt are studied by GSF and ECN (with subcontract to RIVM). The overall objectives of these studies are to develop and demonstrate procedures for the radiological safety assessment of a deep repository in salt formations. An essential objective is to show how far appropriate choices of the repository design parameters can improve the performances of the whole system. The research covers two waste inventories (the Dutch OPLA and the PACOMA reference inventory), two disposal techniques (conventional and solution mining) and three types of formations (salt dome, pillow and bedded salt). An important part of the research has been carried out in the socalled VEOS project within the framework of the Dutch OPLA study. The methodology used in the consequence analysis is a deterministic one. The models and calculation tools used to perform the consequence analysis are the codes: EMOS, METROPOL and BIOS. The results are expressed in terms of dose rates and doses to individuals as well as to groups. Detailed information with respect to the input data and the results obtained with the three codes is given in three annexes to this final report

  6. Coded diffraction system in X-ray crystallography using a boolean phase coded aperture approximation

    Science.gov (United States)

    Pinilla, Samuel; Poveda, Juan; Arguello, Henry

    2018-03-01

    Phase retrieval is a problem present in many applications such as optics, astronomical imaging, computational biology and X-ray crystallography. Recent work has shown that the phase can be better recovered when the acquisition architecture includes a coded aperture, which modulates the signal before diffraction, such that the underlying signal is recovered from coded diffraction patterns. Moreover, this type of modulation effect, before the diffraction operation, can be obtained using a phase coded aperture, just after the sample under study. However, a practical implementation of a phase coded aperture in an X-ray application is not feasible, because it is computationally modeled as a matrix with complex entries which requires changing the phase of the diffracted beams. In fact, changing the phase implies finding a material that allows to deviate the direction of an X-ray beam, which can considerably increase the implementation costs. Hence, this paper describes a low cost coded X-ray diffraction system based on block-unblock coded apertures that enables phase reconstruction. The proposed system approximates the phase coded aperture with a block-unblock coded aperture by using the detour-phase method. Moreover, the SAXS/WAXS X-ray crystallography software was used to simulate the diffraction patterns of a real crystal structure called Rhombic Dodecahedron. Additionally, several simulations were carried out to analyze the performance of block-unblock approximations in recovering the phase, using the simulated diffraction patterns. Furthermore, the quality of the reconstructions was measured in terms of the Peak Signal to Noise Ratio (PSNR). Results show that the performance of the block-unblock phase coded apertures approximation decreases at most 12.5% compared with the phase coded apertures. Moreover, the quality of the reconstructions using the boolean approximations is up to 2.5 dB of PSNR less with respect to the phase coded aperture reconstructions.

  7. SURE: a system of computer codes for performing sensitivity/uncertainty analyses with the RELAP code

    International Nuclear Information System (INIS)

    Bjerke, M.A.

    1983-02-01

    A package of computer codes has been developed to perform a nonlinear uncertainty analysis on transient thermal-hydraulic systems which are modeled with the RELAP computer code. Using an uncertainty around the analyses of experiments in the PWR-BDHT Separate Effects Program at Oak Ridge National Laboratory. The use of FORTRAN programs running interactively on the PDP-10 computer has made the system very easy to use and provided great flexibility in the choice of processing paths. Several experiments simulating a loss-of-coolant accident in a nuclear reactor have been successfully analyzed. It has been shown that the system can be automated easily to further simplify its use and that the conversion of the entire system to a base code other than RELAP is possible

  8. Improved decoding for a concatenated coding system

    DEFF Research Database (Denmark)

    Paaske, Erik

    1990-01-01

    The concatenated coding system recommended by CCSDS (Consultative Committee for Space Data Systems) uses an outer (255,233) Reed-Solomon (RS) code based on 8-b symbols, followed by the block interleaver and an inner rate 1/2 convolutional code with memory 6. Viterbi decoding is assumed. Two new...... decoding procedures based on repeated decoding trials and exchange of information between the two decoders and the deinterleaver are proposed. In the first one, where the improvement is 0.3-0.4 dB, only the RS decoder performs repeated trials. In the second one, where the improvement is 0.5-0.6 dB, both...... decoders perform repeated decoding trials and decoding information is exchanged between them...

  9. Performance Analysis of Optical Code Division Multiplex System

    Science.gov (United States)

    Kaur, Sandeep; Bhatia, Kamaljit Singh

    2013-12-01

    This paper presents the Pseudo-Orthogonal Code generator for Optical Code Division Multiple Access (OCDMA) system which helps to reduce the need of bandwidth expansion and improve spectral efficiency. In this paper we investigate the performance of multi-user OCDMA system to achieve data rate more than 1 Tbit/s.

  10. Bile salt receptor complex activates a pathogenic type III secretion system

    Science.gov (United States)

    Li, Peng; Rivera-Cancel, Giomar; Kinch, Lisa N; Salomon, Dor; Tomchick, Diana R; Grishin, Nick V; Orth, Kim

    2016-01-01

    Bile is an important component of the human gastrointestinal tract with an essential role in food absorption and antimicrobial activities. Enteric bacterial pathogens have developed strategies to sense bile as an environmental cue to regulate virulence genes during infection. We discovered that Vibrio parahaemolyticus VtrC, along with VtrA and VtrB, are required for activating the virulence type III secretion system 2 in response to bile salts. The VtrA/VtrC complex activates VtrB in the presence of bile salts. The crystal structure of the periplasmic domains of the VtrA/VtrC heterodimer reveals a β-barrel with a hydrophobic inner chamber. A co-crystal structure of VtrA/VtrC with bile salt, along with biophysical and mutational analysis, demonstrates that the hydrophobic chamber binds bile salts and activates the virulence network. As part of a family of conserved signaling receptors, VtrA/VtrC provides structural and functional insights into the evolutionarily conserved mechanism used by bacteria to sense their environment. DOI: http://dx.doi.org/10.7554/eLife.15718.001 PMID:27377244

  11. Bile salt receptor complex activates a pathogenic type III secretion system

    Energy Technology Data Exchange (ETDEWEB)

    Li, Peng; Rivera-Cancel, Giomar; Kinch, Lisa N.; Salomon, Dor; Tomchick, Diana R.; Grishin, Nick V.; Orth, Kim

    2016-07-05

    Bile is an important component of the human gastrointestinal tract with an essential role in food absorption and antimicrobial activities. Enteric bacterial pathogens have developed strategies to sense bile as an environmental cue to regulate virulence genes during infection. We discovered thatVibrio parahaemolyticusVtrC, along with VtrA and VtrB, are required for activating the virulence type III secretion system 2 in response to bile salts. The VtrA/VtrC complex activates VtrB in the presence of bile salts. The crystal structure of the periplasmic domains of the VtrA/VtrC heterodimer reveals a β-barrel with a hydrophobic inner chamber. A co-crystal structure of VtrA/VtrC with bile salt, along with biophysical and mutational analysis, demonstrates that the hydrophobic chamber binds bile salts and activates the virulence network. As part of a family of conserved signaling receptors, VtrA/VtrC provides structural and functional insights into the evolutionarily conserved mechanism used by bacteria to sense their environment.

  12. Calculation of density and permeability of compacted crushed salt within an engineered shaft sealing system

    International Nuclear Information System (INIS)

    Loken, M.; Statham, W.

    1997-01-01

    Crushed salt from the host Salado Formation is proposed as a sealing material in one component of a multicomponent seal system design for the shafts of the Waste Isolation Pilot Plant (WIPP), a mined geological repository for storage and disposal of transuranic radioactive wastes located near Carlsbad, New Mexico. The crushed salt will be compacted and placed at a density approaching 90% of the intact density of the host Salado salt. Creep closure of the shaft will further compact the crushed salt over time, thereby reducing the crushed-salt permeability from the initial state and creating an effective long-term seal. A structural model and a fluid flow model have been developed to provide an estimate of crushed-salt reconsolidation rate as a function of depth, time, and pore pressure. Model results are obtained in terms of crushed-salt permeability as a function of time and depth within the salt column. Model results indicate that average salt column permeability will be reduced to 3.3 x 10 -20 m 2 in about 100 years, which provides for an acceptable long-term seal component

  13. Tests of prototype salt stripper system for IFR fuel cycle

    International Nuclear Information System (INIS)

    Carls, E.L.; Blaskovitz, R.J.; Johnson, T.R.; Ogata, T.

    1993-01-01

    One of the waste treatment steps for the on-site reprocessing of spent fuel from the Integral Fast Reactor fuel cycles is stripping of the electrolyte salt used in the electrorefining process. This involves the chemical reduction of the actinides and rare earth chlorides forming metals which then dissolve in a cadmium pool. To develop the equipment for this step, a prototype salt stripper system has been installed in an engineering scale argon-filled glovebox. Pumping trails were successful in transferring 90 kg of LiCl-KCl salt containing uranium and rare earth metal chlorides at 500 degree C from an electrorefiner to the stripper vessel at a pumping rate of about 5 L/min. The freeze seal solder connectors which were used to join sections of the pump and transfer line performed well. Stripping tests have commenced employing an inverted cup charging device to introduce a Cd-15 wt % Li alloy reductant to the stripper vessel

  14. Development of a nuclear power plant system analysis code

    International Nuclear Information System (INIS)

    Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.

    1997-07-01

    During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs

  15. High-Level Waste Salt Disposition Systems Engineering Team Final Report, Volumes I, II, and III

    International Nuclear Information System (INIS)

    Piccolo, S.F.

    1999-01-01

    This report describes the process used and results obtained by the High Level Waste Salt Disposition Systems Engineering Team to select a primary and backup alternative salt disposition method for the Savannah River Site

  16. Software coding for reliable data communication in a reactor safety system

    International Nuclear Information System (INIS)

    Maghsoodi, R.

    1978-01-01

    A software coding method is proposed to improve the communication reliability of a microprocessor based fast-reactor safety system. This method which replaces the conventional coding circuitry, applies a program to code the data which is communicated between the processors via their data memories. The system requirements are studied and the suitable codes are suggested. The problems associated with hardware coders, and the advantages of software coding methods are discussed. The product code which proves a faster coding time over the cyclic code is chosen as the final code. Then the improvement of the communication reliability is derived for a processor and its data memory. The result is used to calculate the reliability improvement of the processing channel as the basic unit for the safety system. (author)

  17. A coupled systems code-CFD MHD solver for fusion blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Wolfendale, Michael J., E-mail: m.wolfendale11@imperial.ac.uk; Bluck, Michael J.

    2015-10-15

    Highlights: • A coupled systems code-CFD MHD solver for fusion blanket applications is proposed. • Development of a thermal hydraulic systems code with MHD capabilities is detailed. • A code coupling methodology based on the use of TCP socket communications is detailed. • Validation cases are briefly discussed for the systems code and coupled solver. - Abstract: The network of flow channels in a fusion blanket can be modelled using a 1D thermal hydraulic systems code. For more complex components such as junctions and manifolds, the simplifications employed in such codes can become invalid, requiring more detailed analyses. For magnetic confinement reactor blanket designs using a conducting fluid as coolant/breeder, the difficulties in flow modelling are particularly severe due to MHD effects. Blanket analysis is an ideal candidate for the application of a code coupling methodology, with a thermal hydraulic systems code modelling portions of the blanket amenable to 1D analysis, and CFD providing detail where necessary. A systems code, MHD-SYS, has been developed and validated against existing analyses. The code shows good agreement in the prediction of MHD pressure loss and the temperature profile in the fluid and wall regions of the blanket breeding zone. MHD-SYS has been coupled to an MHD solver developed in OpenFOAM and the coupled solver validated for test geometries in preparation for modelling blanket systems.

  18. Uncertainty and sensitivity analysis using probabilistic system assessment code. 1

    International Nuclear Information System (INIS)

    Honma, Toshimitsu; Sasahara, Takashi.

    1993-10-01

    This report presents the results obtained when applying the probabilistic system assessment code under development to the PSACOIN Level 0 intercomparison exercise organized by the Probabilistic System Assessment Code User Group in the Nuclear Energy Agency (NEA) of OECD. This exercise is one of a series designed to compare and verify probabilistic codes in the performance assessment of geological radioactive waste disposal facilities. The computations were performed using the Monte Carlo sampling code PREP and post-processor code USAMO. The submodels in the waste disposal system were described and coded with the specification of the exercise. Besides the results required for the exercise, further additional uncertainty and sensitivity analyses were performed and the details of these are also included. (author)

  19. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

    Suyama, Kenya; Hirakawa, Naohiro; Iwasaki, Tomohiko.

    1997-11-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)

  20. Development and verification of a coupled code system RETRAN-MASTER-TORC

    International Nuclear Information System (INIS)

    Cho, J.Y.; Song, J.S.; Joo, H.G.; Zee, S.Q.

    2004-01-01

    Recently, coupled thermal-hydraulics (T-H) and three-dimensional kinetics codes have been widely used for the best-estimate simulations such as the main steam line break (MSLB) and locked rotor problems. This work is to develop and verify one of such codes by coupling the system T-H code RETRAN, the 3-D kinetics code MASTER and sub-channel analysis code TORC. The MASTER code has already been applied to such simulations after coupling with the MARS or RETRAN-3D multi-dimensional system T-H codes. The MASTER code contains a sub-channel analysis code COBRA-III C/P, and the coupled systems MARSMASTER-COBRA and RETRAN-MASTER-COBRA had been already developed and verified. With these previous studies, a new coupled system of RETRAN-MASTER-TORC is to be developed and verified for the standard best-estimate simulation code package in Korea. The TORC code has already been applied to the thermal hydraulics design of the several ABB/CE type plants and Korean Standard Nuclear Power Plants (KSNP). This justifies the choice of TORC rather than COBRA. Because the coupling between RETRAN and MASTER codes are already established and verified, this work is simplified to couple the TORC sub-channel T-H code with the MASTER neutronics code. The TORC code is a standalone code that solves the T-H equations for a given core problem from reading the input file and finally printing the converged solutions. However, in the coupled system, because TORC receives the pin power distributions from the neutronics code MASTER and transfers the T-H results to MASTER iteratively, TORC needs to be controlled by the MASTER code and does not need to solve the given problem completely at each iteration step. By this reason, the coupling of the TORC code with the MASTER code requires several modifications in the I/O treatment, flow iteration and calculation logics. The next section of this paper describes the modifications in the TORC code. The TORC control logic of the MASTER code is then followed. The

  1. Advanced video coding systems

    CERN Document Server

    Gao, Wen

    2015-01-01

    This comprehensive and accessible text/reference presents an overview of the state of the art in video coding technology. Specifically, the book introduces the tools of the AVS2 standard, describing how AVS2 can help to achieve a significant improvement in coding efficiency for future video networks and applications by incorporating smarter coding tools such as scene video coding. Topics and features: introduces the basic concepts in video coding, and presents a short history of video coding technology and standards; reviews the coding framework, main coding tools, and syntax structure of AV

  2. Evaluation of the analysis models in the ASTRA nuclear design code system

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Park, Chang Jea; Kim, Do Sam; Lee, Kyeong Taek; Kim, Jong Woon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-11-15

    In the field of nuclear reactor design, main practice was the application of the improved design code systems. During the process, a lot of basis and knowledge were accumulated in processing input data, nuclear fuel reload design, production and analysis of design data, et al. However less efforts were done in the analysis of the methodology and in the development or improvement of those code systems. Recently, KEPO Nuclear Fuel Company (KNFC) developed the ASTRA (Advanced Static and Transient Reactor Analyzer) code system for the purpose of nuclear reactor design and analysis. In the code system, two group constants were generated from the CASMO-3 code system. The objective of this research is to analyze the analysis models used in the ASTRA/CASMO-3 code system. This evaluation requires indepth comprehension of the models, which is important so much as the development of the code system itself. Currently, most of the code systems used in domestic Nuclear Power Plant were imported, so it is very difficult to maintain and treat the change of the situation in the system. Therefore, the evaluation of analysis models in the ASTRA nuclear reactor design code system in very important.

  3. High-temperature molten salt thermal energy storage systems for solar applications

    Science.gov (United States)

    Petri, R. J.; Claar, T. D.; Ong, E.

    1983-01-01

    Experimental results of compatibility screening studies of 100 salt/containment/thermal conductivity enhancement (TCE) combinations for the high temperature solar thermal application range of 704 deg to 871 C (1300 to 1600 F) are presented. Nine candidate containment/HX alloy materials and two TCE materials were tested with six candidate solar thermal alkali and alkaline earth carbonate storage salts (both reagent and technical grade of each). Compatibility tests were conducted with salt encapsulated in approx. 6.0 inch x 1 inch welded containers of test material from 300 to 3000 hours. Compatibility evaluations were end application oriented, considering the potential 30 year lifetime requirement of solar thermal power plant components. Analyses were based on depth and nature of salt side corrosion of materials, containment alloy thermal aging effects, weld integrity in salt environment, air side containment oxidation, and chemical and physical analyses of the salt. A need for more reliable, and in some cases first time determined thermophysical and transport property data was also identified for molten carbonates in the 704 to 871 C temperature range. In particular, accurate melting point (mp) measurements were performed for Li2CO3 and Na2CO3 while melting point, heat of fusion, and specific heat determinations were conducted on 81.3 weight percent Na2CO3-18.7 weight percent K2CO3 and 52.2 weight percent BaCO3-47.8 weight percent Na2CO3 to support future TES system design and ultimate scale up of solar thermal energy storage (TES) subsystems.

  4. Data exchange between zero dimensional code and physics platform in the CFETR integrated system code

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Guoliang [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Shi, Nan [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Zhou, Yifu; Mao, Shifeng [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Jian, Xiang [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronics Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Chen, Jiale [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Liu, Li; Chan, Vincent [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Ye, Minyou, E-mail: yemy@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China)

    2016-11-01

    Highlights: • The workflow of the zero dimensional code and the multi-dimension physics platform of CFETR integrated system codeis introduced. • The iteration process among the codes in the physics platform. • The data transfer between the zero dimensionalcode and the physical platform, including data iteration and validation, and justification for performance parameters.. - Abstract: The China Fusion Engineering Test Reactor (CFETR) integrated system code contains three parts: a zero dimensional code, a physics platform and an engineering platform. We use the zero dimensional code to identify a set of preliminary physics and engineering parameters for CFETR, which is used as input to initiate multi-dimension studies using the physics and engineering platform for design, verification and validation. Effective data exchange between the zero dimensional code and the physical platform is critical for the optimization of CFETR design. For example, in evaluating the impact of impurity radiation on core performance, an open field line code is used to calculate the impurity transport from the first-wall boundary to the pedestal. The impurity particle in the pedestal are used as boundary conditions in a transport code for calculating impurity transport in the core plasma and the impact of core radiation on core performance. Comparison of the results from the multi-dimensional study to those from the zero dimensional code is used to further refine the controlled radiation model. The data transfer between the zero dimensional code and the physical platform, including data iteration and validation, and justification for performance parameters will be presented in this paper.

  5. A study on the nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

    Kim, Yeon Seung; Huh, Young Hwan; Lee, Jong Bok; Choi, Young Gil; Suh, Soong Hyok; Kang, Byong Heon; Kim, Hee Kyung; Kim, Ko Ryeo; Park, Soo Jin

    1990-12-01

    According to current software development and quality assurance trends. It is necessary to develop computer code management system for nuclear programs. For this reason, the project started in 1987. Main objectives of the project are to establish a nuclear computer code management system, to secure software reliability, and to develop nuclear computer code packages. Contents of performing the project in this year were to operate and maintain computer code information system of KAERI computer codes, to develop application tool, AUTO-i, for solving the 1st and 2nd moments of inertia on polygon or circle, and to research nuclear computer code conversion between different machines. For better supporting the nuclear code availability and reliability, assistance from users who are using codes is required. Lastly, for easy reference about the codes information, we presented list of code names and information on the codes which were introduced or developed during this year. (Author)

  6. Design of ACM system based on non-greedy punctured LDPC codes

    Science.gov (United States)

    Lu, Zijun; Jiang, Zihong; Zhou, Lin; He, Yucheng

    2017-08-01

    In this paper, an adaptive coded modulation (ACM) scheme based on rate-compatible LDPC (RC-LDPC) codes was designed. The RC-LDPC codes were constructed by a non-greedy puncturing method which showed good performance in high code rate region. Moreover, the incremental redundancy scheme of LDPC-based ACM system over AWGN channel was proposed. By this scheme, code rates vary from 2/3 to 5/6 and the complication of the ACM system is lowered. Simulations show that more and more obvious coding gain can be obtained by the proposed ACM system with higher throughput.

  7. Application of salting-out effect equation to modelling of liquid-liquid distribution systems

    International Nuclear Information System (INIS)

    Pitsch, H.K.

    1986-03-01

    Physicochemical interpretation of salting-out is reviewed and effects of the medium on liquid-liquid distribution equilibria are described by two non-specific parameters of salting-out agents: total concentration of species in the aqueous phase and water activity. Thus extraction of a given constituent in various media can be forecasted with few data. Different uranyl and technetium (VII) extraction systems are analyzed to show the potentiality of the method. Coextraction of nitric acid and uranyl nitrate by tributyl phosphate is used to show the possibility of modelling complex distribution systems in industrial conditions [fr

  8. Implementing a modular system of computer codes

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1983-07-01

    A modular computation system has been developed for nuclear reactor core analysis. The codes can be applied repeatedly in blocks without extensive user input data, as needed for reactor history calculations. The primary control options over the calculational paths and task assignments within the codes are blocked separately from other instructions, admitting ready access by user input instruction or directions from automated procedures and promoting flexible and diverse applications at minimum application cost. Data interfacing is done under formal specifications with data files manipulated by an informed manager. This report emphasizes the system aspects and the development of useful capability, hopefully informative and useful to anyone developing a modular code system of much sophistication. Overall, this report in a general way summarizes the many factors and difficulties that are faced in making reactor core calculations, based on the experience of the authors. It provides the background on which work on HTGR reactor physics is being carried out

  9. Code-modulated interferometric imaging system using phased arrays

    Science.gov (United States)

    Chauhan, Vikas; Greene, Kevin; Floyd, Brian

    2016-05-01

    Millimeter-wave (mm-wave) imaging provides compelling capabilities for security screening, navigation, and bio- medical applications. Traditional scanned or focal-plane mm-wave imagers are bulky and costly. In contrast, phased-array hardware developed for mass-market wireless communications and automotive radar promise to be extremely low cost. In this work, we present techniques which can allow low-cost phased-array receivers to be reconfigured or re-purposed as interferometric imagers, removing the need for custom hardware and thereby reducing cost. Since traditional phased arrays power combine incoming signals prior to digitization, orthogonal code-modulation is applied to each incoming signal using phase shifters within each front-end and two-bit codes. These code-modulated signals can then be combined and processed coherently through a shared hardware path. Once digitized, visibility functions can be recovered through squaring and code-demultiplexing operations. Pro- vided that codes are selected such that the product of two orthogonal codes is a third unique and orthogonal code, it is possible to demultiplex complex visibility functions directly. As such, the proposed system modulates incoming signals but demodulates desired correlations. In this work, we present the operation of the system, a validation of its operation using behavioral models of a traditional phased array, and a benchmarking of the code-modulated interferometer against traditional interferometer and focal-plane arrays.

  10. Diffusion Welding of Alloys for Molten Salt Service - Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Denis Clark; Ronald Mizia; Piyush Sabharwall

    2012-09-01

    The present work is concerned with heat exchanger development for molten salt service, including the proposed molten salt reactor (MSR), a homogeneous reactor in which the fuel is dissolved in a circulating fluid of molten salt. It is an outgrowth of recent work done under the Next Generation Nuclear Plant (NGNP) program; what the two reactor systems have in common is an inherently safe nuclear plant with a high outlet temperature that is useful for process heat as well as more conventional generation The NGNP program was tasked with investigating the application of a new generation of nuclear power plants to a variety of energy needs. One baseline reactor design for this program is a high temperature, gas-cooled reactor (HTGR), which provides many options for energy use. These might include the conventional Rankine cycle (steam turbine) generation of electricity, but also other methods: for example, Brayton cycle (gas turbine) electrical generation, and the direct use of the high temperatures characteristic of HTGR output for process heat in the chemical industry. Such process heat is currently generated by burning fossil fuels, and is a major contributor to the carbon footprint of the chemical and petrochemical industries. The HTGR, based on graphite fuel elements, can produce very high output temperatures; ideally, temperatures of 900 °C or even greater, which has significant energy advantages. Such temperatures are, of course, at the frontiers of materials limitations, at the upper end of the performance envelope of the metallic materials for which robust construction codes exist, and within the realm of ceramic materials, the fabrication and joining of which, on the scale of large energy systems, are at an earlier stage of development. A considerable amount of work was done in the diffusion welding of materials of interest for HTGR service with alloys such as 617 and 800H. The MSR output temperature is also materials limited, and is projected at about 700

  11. TOUGH+HYDRATE v1.2 User's Manual: A Code for the Simulation of System Behavior in Hydrate-Bearing Geologic Media

    Energy Technology Data Exchange (ETDEWEB)

    Moridis, George J. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Kowalsky, Michael B. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Pruess, Karsten [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2012-08-01

    TOUGH+HYDRATE v1.2 is a code for the simulation of the behavior of hydratebearing geologic systems, and represents the second update of the code since its first release [Moridis et al., 2008]. By solving the coupled equations of mass and heat balance, TOUGH+HYDRATE can model the non-isothermal gas release, phase behavior and flow of fluids and heat under conditions typical of common natural CH4-hydrate deposits (i.e., in the permafrost and in deep ocean sediments) in complex geological media at any scale (from laboratory to reservoir) at which Darcy’s law is valid. TOUGH+HYDRATE v1.2 includes both an equilibrium and a kinetic model of hydrate formation and dissociation. The model accounts for heat and up to four mass components, i.e., water, CH4, hydrate, and water-soluble inhibitors such as salts or alcohols. These are partitioned among four possible phases (gas phase, liquid phase, ice phase and hydrate phase). Hydrate dissociation or formation, phase changes and the corresponding thermal effects are fully described, as are the effects of inhibitors. The model can describe all possible hydrate dissociation mechanisms, i.e., depressurization, thermal stimulation, salting-out effects and inhibitor-induced effects. TOUGH+HYDRATE is a member of TOUGH+, the successor to the TOUGH2 [Pruess et al., 1991] family of codes for multi-component, multiphase fluid and heat flow developed at the Lawrence Berkeley National Laboratory. It is written in standard FORTRAN 95/2003, and can be run on any computational platform (workstation, PC, Macintosh) for which such compilers are available.

  12. Optical code division multiple access secure communications systems with rapid reconfigurable polarization shift key user code

    Science.gov (United States)

    Gao, Kaiqiang; Wu, Chongqing; Sheng, Xinzhi; Shang, Chao; Liu, Lanlan; Wang, Jian

    2015-09-01

    An optical code division multiple access (OCDMA) secure communications system scheme with rapid reconfigurable polarization shift key (Pol-SK) bipolar user code is proposed and demonstrated. Compared to fix code OCDMA, by constantly changing the user code, the performance of anti-eavesdropping is greatly improved. The Pol-SK OCDMA experiment with a 10 Gchip/s user code and a 1.25 Gb/s user data of payload has been realized, which means this scheme has better tolerance and could be easily realized.

  13. Near-Capacity Coding for Discrete Multitone Systems with Impulse Noise

    Directory of Open Access Journals (Sweden)

    Kschischang Frank R

    2006-01-01

    Full Text Available We consider the design of near-capacity-achieving error-correcting codes for a discrete multitone (DMT system in the presence of both additive white Gaussian noise and impulse noise. Impulse noise is one of the main channel impairments for digital subscriber lines (DSL. One way to combat impulse noise is to detect the presence of the impulses and to declare an erasure when an impulse occurs. In this paper, we propose a coding system based on low-density parity-check (LDPC codes and bit-interleaved coded modulation that is capable of taking advantage of the knowledge of erasures. We show that by carefully choosing the degree distribution of an irregular LDPC code, both the additive noise and the erasures can be handled by a single code, thus eliminating the need for an outer code. Such a system can perform close to the capacity of the channel and for the same redundancy is significantly more immune to the impulse noise than existing methods based on an outer Reed-Solomon (RS code. The proposed method has a lower implementation complexity than the concatenated coding approach.

  14. The molten salt reactors (MSR) pyro chemistry and fuel cycle for innovative nuclear systems

    International Nuclear Information System (INIS)

    Brossard, Ph.; Garzenne, C.; Mouney, H.

    2002-01-01

    In the frame of the studies on next generation nuclear systems, and especially for the molten salt reactors and for the integrated fuel cycle (as IFR), the fuel cycle constraints must be taken into account in the preliminary studies of the system to improve the cycle and reactor optimisation. Among the purposes for next generation nuclear systems, sustainability and waste (radio-toxicity and mass) management are important goals. These goals imply reprocessing and recycling strategies. The objectives of this workshop are to present and to share the different strategies and scenarios, the needs based on these scenarios, the experimental facilities available today or in the future and their capabilities, the needs for demonstration. It aims at: identifying the needs for fuel cycle based on solid fuel or liquid fuel, and especially, the on-line reprocessing or clean up for the molten salt reactors; assessing the state-of-the-art on the pyro-chemistry applied to solid fuel and to present the research activities; assessing the state-of-the-art on liquid fuels (or others), and to present the research activities; expressing the R and D programs for pyro-chemistry, molten salt, and also to propose innovative processes; and proposing some joint activities in the frame of GEDEON and PRACTIS programs. This document brings together the transparencies of 18 contributions dealing with: scenario studies with AMSTER concept (Scenarios, MSR, breeders (Th) and burners); fuel cycle for innovative systems; current reprocessing of spent nuclear fuel (SNF) in molten salts (review of pyro-chemistry processes (non nuclear and nuclear)); high temperature NMR spectroscopies in molten salts; reductive extraction of An from molten fluorides (salt - liquid metal extraction); electrochemistry characterisation; characterisation with physical methods - extraction coefficient and kinetics; electrolytic extraction; dissolution-precipitation of plutonium in the eutectic LiCl-KCl (dissolution and

  15. A Systematic Review of Coding Systems Used in Pharmacoepidemiology and Database Research.

    Science.gov (United States)

    Chen, Yong; Zivkovic, Marko; Wang, Tongtong; Su, Su; Lee, Jianyi; Bortnichak, Edward A

    2018-02-01

    Clinical coding systems have been developed to translate real-world healthcare information such as prescriptions, diagnoses and procedures into standardized codes appropriate for use in large healthcare datasets. Due to the lack of information on coding system characteristics and insufficient uniformity in coding practices, there is a growing need for better understanding of coding systems and their use in pharmacoepidemiology and observational real world data research. To determine: 1) the number of available coding systems and their characteristics, 2) which pharmacoepidemiology databases are they adopted in, 3) what outcomes and exposures can be identified from each coding system, and 4) how robust they are with respect to consistency and validity in pharmacoepidemiology and observational database studies. Electronic literature database and unpublished literature searches, as well as hand searching of relevant journals were conducted to identify eligible articles discussing characteristics and applications of coding systems in use and published in the English language between 1986 and 2016. Characteristics considered included type of information captured by codes, clinical setting(s) of use, adoption by a pharmacoepidemiology database, region, and available mappings. Applications articles describing the use and validity of specific codes, code lists, or algorithms were also included. Data extraction was performed independently by two reviewers and a narrative synthesis was performed. A total of 897 unique articles and 57 coding systems were identified, 17% of which included country-specific modifications or multiple versions. Procedures (55%), diagnoses (36%), drugs (38%), and site of disease (39%) were most commonly and directly captured by these coding systems. The systems were used to capture information from the following clinical settings: inpatient (63%), ambulatory (55%), emergency department (ED, 34%), and pharmacy (13%). More than half of all coding

  16. Geometrical Approach to the Grid System in the KOPEC Pilot Code

    International Nuclear Information System (INIS)

    Lee, E. J.; Park, C. E.; Lee, S. Y.

    2008-01-01

    KOPEC has been developing a pilot code to analyze two phase flow. The earlier version of the pilot code adopts the geometry with one-dimensional structured mesh system. As the pilot code is required to handle more complex geometries, a systematic geometrical approach to grid system has been introduced. Grid system can be classified as two types; structured grid system and unstructured grid system. The structured grid system is simple to apply but is less flexible than the other. The unstructured grid system is more complicated than the structured grid system. But it is more flexible to model the geometry. Therefore, two types of grid systems are utilized to allow code users simplicity as well as the flexibility

  17. Development of System Based Code: Case Study of Life-Cycle Margin Evaluation

    International Nuclear Information System (INIS)

    Tai Asayama; Masaki Morishita; Masanori Tashimo

    2006-01-01

    For a leap of progress in structural deign of nuclear plant components, The late Professor Emeritus Yasuhide Asada proposed the System Based Code. The key concepts of the System Based Code are; (1) life-cycle margin optimization, (2) expansion of technical options as well as combinations of technical options beyond the current codes and standards, and (3) designing to clearly defined target reliabilities. Those concepts are very new to most of the nuclear power plant designers who are naturally obliged to design to current codes and standards; the application of the concepts of the System Based Code to design will lead to entire change of practices that designers have long been accustomed to. On the other hand, experienced designers are supposed to have expertise that can support and accelerate the development of the System Based Code. Therefore, interfacing with experienced designers is of crucial importance for the development of the System Based Code. The authors conducted a survey on the acceptability of the System Based Code concept. The results were analyzed from the possibility of improving structural design both in terms of reliability and cost effectiveness by the introduction of the System Based Code concept. It was concluded that the System Based Code is beneficial for those purposes. Also described is the expertise elicited from the results of the survey that can be reflected to the development of the System Based Code. (authors)

  18. Summary description of the scale modular code system

    International Nuclear Information System (INIS)

    Parks, C.V.

    1987-12-01

    SCALE - a modular code system for Standardized Computer Analyses for Licensing Evaluation - has been developed at Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission staff. The SCALE system utilizes well-established computer codes and methods within standard analytic sequences that allow simplified free-form input, automate the data processing and coupling between codes, and provide accurate and reliable results. System development has been directed at criticality safety, shielding, and heat transfer analysis of spent fuel transport and/or storage casks. However, only a few of the sequences (and none of the individual functional modules) are restricted to cask applications. This report will provide a background on the history of the SCALE development and review the components and their function within the system. The available data libraries are also discussed, together with the automated features that standardize the data processing and systems analysis. 83 refs., 32 figs., 11 tabs

  19. Speciation of Zinc Mixed Ligand Complexes in Salt Water Systems ...

    African Journals Online (AJOL)

    Speciation of Zinc Mixed Ligand Complexes in Salt Water Systems. ... method has been used to study heavy metal interaction in model lake water in KNO3 ... is of no consequential effect because in its normal state, the [OH-] of the lake water is ...

  20. Modular Modeling System (MMS) code: a versatile power plant analysis package

    International Nuclear Information System (INIS)

    Divakaruni, S.M.; Wong, F.K.L.

    1987-01-01

    The basic version of the Modular Modeling System (MMS-01), a power plant systems analysis computer code jointly developed by the Nuclear Power and the Coal Combustion Systems Divisions of the Electric Power Research Institute (EPRI), has been released to the utility power industry in April 1983 at a code release workshop held in Charlotte, North Carolina. Since then, additional modules have been developed to analyze the Pressurized Water Reactors (PWRs) and the Boiling Water Reactors (BWRs) when the safety systems are activated. Also, a selected number of modules in the MMS-01 library have been modified to allow the code users more flexibility in constructing plant specific systems for analysis. These new PWR and BWR modules constitute the new MMS library, and it includes the modifications to the MMS-01 library. A year and half long extensive code qualification program of this new version of the MMS code at EPRI and the contractor sites, back by further code testing in an user group environment is culminating in the MMS-02 code release announcement seminar. At this seminar, the results of user group efforts and the code qualification program will be presented in a series of technical sessions. A total of forty-nine papers will be presented to describe the new code features and the code qualification efforts. For the sake of completion, an overview of the code is presented to include the history of the code development, description of the MMS code and its structure, utility engineers involvement in MMS-01 and MMS-02 validations, the enhancements made in the last 18 months to the code, and finally the perspective on the code future in the fossil and nuclear industry

  1. Simulating Salt Movement using a Coupled Salinity Transport Model in a Variably Saturated Agricultural Groundwater System

    Science.gov (United States)

    Tavakoli Kivi, S.; Bailey, R. T.; Gates, T. K.

    2017-12-01

    Salinization is one of the major concerns in irrigated agricultural fields. Increasing salinity concentrations are due principally to a high water table that results from excessive irrigation, canal seepage, and a lack of efficient drainage systems, and lead to decreasing crop yield. High groundwater salinity loading to nearby river systems also impacts downstream areas, with saline river water diverted for application on irrigated fields. To assess the different strategies for salt remediation, we present a reactive transport model (UZF-RT3D) coupled with a salinity equilibrium chemistry module for simulating the fate and transport of salt ions in a variably-saturated agricultural groundwater system. The developed model accounts not for advection, dispersion, nitrogen and sulfur cycling, oxidation-reduction, sorption, complexation, ion exchange, and precipitation/dissolution of salt minerals. The model is applied to a 500 km2 region within the Lower Arkansas River Valley (LARV) in southeastern Colorado, an area acutely affected by salinization in the past few decades. The model is tested against salt ion concentrations in the saturated zone, total dissolved solid concentrations in the unsaturated zone, and salt groundwater loading to the Arkansas River. The model now can be used to investigate salinity remediation strategies.

  2. A calculational procedure for neutronic and depletion analysis of Molten-Salt reactors based on SCALE6/TRITON

    International Nuclear Information System (INIS)

    Sheu, R.J.; Chang, J.S.; Liu, Y.-W. H.

    2011-01-01

    Molten-Salt Reactors (MSRs) represent one of the selected categories in the GEN-IV program. This type of reactor is distinguished by the use of liquid fuel circulating in and out of the core, which makes it possible for online refueling and salt processing. However, this operation characteristic also complicates the modeling and simulation of reactor core behaviour using conventional neutronic codes. The TRITON sequence in the SCALE6 code system has been designed to provide the combined capabilities of problem-dependent cross-section processing, rigorous treatment of neutron transport, and coupled with the ORIGEN-S depletion calculations. In order to accommodate the simulation of dynamic refueling and processing scheme, an in-house program REFRESH together with a run script are developed for carrying out a series of stepwise TRITON calculations, that makes the work of analyzing the neutronic properties and performance of a MSR core design easier. As a demonstration and cross check, we have applied this method to reexamine the conceptual design of Molten Salt Actinide Recycler & Transmuter (MOSART). This paper summarizes the development of the method and preliminary results of its application on MOSART. (author)

  3. Molten salt reactor technology for long-range and wide-scale nuclear energy system

    International Nuclear Information System (INIS)

    Ignatiev, V.; Alexseev, P.; Menshikov, L.; Prusakov, V.; Subbotine, S.

    1997-01-01

    A possibility of creation of multi-component nuclear power system in which alongside with thermal and fast reactors, molten salt burner reactors, for incineration of weapon grade plutonium, some minor actinides and transmutation of some fission products will be presented. The purposes of this work are to review the present status of the molten salt reactor technology and innovative non-aqueous chemical processing methods, to indicate the importance of the uncertainties remaining, to identify the additional work needed, and to evaluate the probability of success in obtaining improved safety characteristics for new concept of molten salt - burner reactor with external neutron source. 8 refs., 3 figs., 2 tabs

  4. Nonterminals and codings in defining variations of OL-systems

    DEFF Research Database (Denmark)

    Skyum, Sven

    1974-01-01

    The use of nonterminals versus the use of codings in variations of OL-systems is studied. It is shown that the use of nonterminals produces a comparatively low generative capacity in deterministic systems while it produces a comparatively high generative capacity in nondeterministic systems. Fina....... Finally it is proved that the family of context-free languages is contained in the family generated by codings on propagating OL-systems with a finite set of axioms, which was one of the open problems in [10]. All the results in this paper can be found in [71] and [72].......The use of nonterminals versus the use of codings in variations of OL-systems is studied. It is shown that the use of nonterminals produces a comparatively low generative capacity in deterministic systems while it produces a comparatively high generative capacity in nondeterministic systems...

  5. Recent developments in the Los Alamos radiation transport code system

    International Nuclear Information System (INIS)

    Forster, R.A.; Parsons, K.

    1997-01-01

    A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results

  6. Salt repository sealing materials development program: 5-year work plan

    International Nuclear Information System (INIS)

    Myers, L.B.

    1986-06-01

    This plan covers 5 years (fiscal years 1986 through 1990) of work in the repository sealing materials program to support design decisions and licensing activities for a salt repository. The plan covers a development activity, not a research activity. There are firm deliverables as the end points of each part of the work. The major deliverables are: development plans for code development and materials testing; seal system components models; seal system performance specifications; seal materials specifications; and seal materials properties ''handbook.'' The work described in this plan is divided into three general tasks as follows: mathematical modeling; materials studies (salt, cementitious materials, and earthen materials); and large-scale testing. Each of the sections presents an overview, status, planned activities, and summary of program milestones. This plan will be the starting point for preparing the development plans described above, but is subject to change if preparation of the work plan indicates that a different approach or sequence is preferable to achieve the ultimate goal, i.e., support of design and licensing

  7. Salt stress-induced changes in antioxidative defense system and proteome profiles of salt-tolerant and sensitive Frankia strains.

    Science.gov (United States)

    Srivastava, Amrita; Singh, Anumeha; Singh, Satya S; Mishra, Arun K

    2017-04-16

    An appreciation of comparative microbial survival is most easily done while evaluating their adaptive strategies during stress. In the present experiment, antioxidative and whole cell proteome variations based on spectrophotometric analysis and SDS-PAGE and 2-dimensional gel electrophoresis have been analysed among salt-tolerant and salt-sensitive Frankia strains. This is the first report of proteomic basis underlying salt tolerance in these newly isolated Frankia strains from Hippophae salicifolia D. Don. Salt-tolerant strain HsIi10 shows higher increment in the contents of superoxide dismutase, catalase and ascorbate peroxidase as compared to salt-sensitive strain HsIi8. Differential 2-DGE profile has revealed differential profiles for salt-tolerant and salt-sensitive strains. Proteomic confirmation of salt tolerance in the strains with inbuilt efficiency of thriving in nitrogen-deficient locales is a definite advantage for these microbes. This would be equally beneficial for improvement of soil nitrogen status. Efficient protein regulation in HsIi10 suggests further exploration for its potential use as biofertilizer in saline soils.

  8. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  9. HTR core physics and transient analyses by the Panthermix code system

    International Nuclear Information System (INIS)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

    2005-01-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

  10. Free-space optical code-division multiple-access system design

    Science.gov (United States)

    Jeromin, Lori L.; Kaufmann, John E.; Bucher, Edward A.

    1993-08-01

    This paper describes an optical direct-detection multiple access communications system for free-space satellite networks utilizing code-division multiple-access (CDMA) and forward error correction (FEC) coding. System performance is characterized by how many simultaneous users operating at data rate R can be accommodated in a signaling bandwidth W. The performance of two CDMA schemes, optical orthogonal codes (OOC) with FEC and orthogonal convolutional codes (OCC), is calculated and compared to information-theoretic capacity bounds. The calculations include the effects of background and detector noise as well as nonzero transmitter extinction ratio and power imbalance among users. A system design for 10 kbps multiple-access communications between low-earth orbit satellites is given. With near- term receiver technology and representative system losses, a 15 W peak-power transmitter provides 10-6 BER performance with seven interfering users and full moon background in the receiver FOV. The receiver employs an array of discrete wide-area avalanche photodiodes (APD) for wide field of view coverage. Issues of user acquisition and synchronization, implementation technology, and system scalability are also discussed.

  11. A comparison of thermal algorithms of fuel rod performance code systems

    International Nuclear Information System (INIS)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

  12. A comparison of thermal algorithms of fuel rod performance code systems

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

  13. Gases in molten salts

    CERN Document Server

    Tomkins, RPT

    1991-01-01

    This volume contains tabulated collections and critical evaluations of original data for the solubility of gases in molten salts, gathered from chemical literature through to the end of 1989. Within the volume, material is arranged according to the individual gas. The gases include hydrogen halides, inert gases, oxygen, nitrogen, hydrogen, carbon dioxide, water vapor and halogens. The molten salts consist of single salts, binary mixtures and multicomponent systems. Included also, is a special section on the solubility of gases in molten silicate systems, focussing on slags and fluxes.

  14. Thermodynamic modelling of phase equilibrium for water + poly(Ethylene glycol + salt aqueous two-phase systems

    Directory of Open Access Journals (Sweden)

    R.A.G. Sé

    2002-04-01

    Full Text Available The NRTL (nonrandom, two-liquid model, expressed in mass fraction instead of mole fraction, was used to correlate liquid-liquid equilibria for aqueous two-phase polymer-salt solutions. New interaction energy parameters for this model were determined using reported data on the water + poly(ethylene glycol + salt systems, with different molecular masses for PEG and the salts potassium phosphate, sodium sulfate, sodium carbonate and magnesium sulfate. The correlation of liquid-liquid equilibrium is quite satisfactory.

  15. Evaluation and implementation of QR Code Identity Tag system for Healthcare in Turkey.

    Science.gov (United States)

    Uzun, Vassilya; Bilgin, Sami

    2016-01-01

    For this study, we designed a QR Code Identity Tag system to integrate into the Turkish healthcare system. This system provides QR code-based medical identification alerts and an in-hospital patient identification system. Every member of the medical system is assigned a unique QR Code Tag; to facilitate medical identification alerts, the QR Code Identity Tag can be worn as a bracelet or necklace or carried as an ID card. Patients must always possess the QR Code Identity bracelets within hospital grounds. These QR code bracelets link to the QR Code Identity website, where detailed information is stored; a smartphone or standalone QR code scanner can be used to scan the code. The design of this system allows authorized personnel (e.g., paramedics, firefighters, or police) to access more detailed patient information than the average smartphone user: emergency service professionals are authorized to access patient medical histories to improve the accuracy of medical treatment. In Istanbul, we tested the self-designed system with 174 participants. To analyze the QR Code Identity Tag system's usability, the participants completed the System Usability Scale questionnaire after using the system.

  16. Supramolecular Complexes Formed in Systems Bile Salt-Bilirubin-Silica

    Science.gov (United States)

    Vlasova, N. N.; Severinovskaya, O. V.; Golovkova, L. P.

    The formation of supramolecular complexes between bilirubin and primary micelles of bile salts has been studied. The association constants of bile salts and binding of bilirubin with these associates have been determined. The adsorption of bilirubin and bile salts from individual and mixed aqueous solutions onto hydrophobic silica surfaces has been investigated. The interaction of bilirubin with primary bile salt micelles and the strong retention in mixed micelles, which are supramolecular complexes, result in the adsorption of bilirubin in free state only.

  17. Engineered Option Treatment of Remediated Nitrate Salts: Surrogate Batch-Blending Testing

    Energy Technology Data Exchange (ETDEWEB)

    Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-11

    This report provides results from batch-blending test work for remediated nitrate salt (RNS) treatment. Batch blending was identified as a preferred option for blending RNS and unremediated nitrate salt (UNS) material with zeolite to effectively safe the salt/Swheat material identified as ignitable (U.S. Environmental Protection Agency code D001). Blending with zeolite was the preferred remediation option identified in the Options Assessment Report and was originally proposed as the best option for remediation by Clark and Funk in their report, Chemical Reactivity and Recommended Remediation Strategy for Los Alamos Remediated Nitrate Salt (RNS) Wastes, and also found to be a preferred option in the Engineering Options Assessment Report: Nitrate Salt Waste Stream Processing. This test work evaluated equipment and recipe alternatives to achieve effective blending of surrogate waste with zeolite.

  18. Where Does Road Salt Go - a Static Salt Model

    Science.gov (United States)

    Yu, C. W.; Liu, F.; Moriarty, V. W.

    2017-12-01

    Each winter, more than 15 million tons of road salt is applied in the United States for the de-icing purpose. Considerable amount of chloride in road salt flows into streams/drainage systems with the snow melt runoff and spring storms, and eventually goes into ecologically sensitive low-lying areas in the watershed, such as ponds and lakes. In many watersheds in the northern part of US, the chloride level in the water body has increased significantly in the past decades, and continues an upward trend. The environmental and ecological impact of the elevated chloride level can no longer be ignored. However although there are many studies on the biological impact of elevated chloride levels, there are few investigations on how the spatially distributed road salt application affects various parts of the watershed. In this presentation, we propose a static road salt model as a first-order metric to address spacial distribution of salt loading. Derived from the Topological Wetness Index (TWI) in many hydrological models, this static salt model provides a spatial impact as- sessment of road salt applications. To demonstrate the effectiveness of the static model, National Elevation Dataset (NED) of ten-meter resolution of Lake George watershed in New York State is used to generate the TWI, which is used to compute a spatially dis- tributed "salt-loading coefficient" of the whole watershed. Spatially varying salt applica- tion rate is then aggregated, using the salt-loading coefficients as weights, to provide salt loading assessments of streams in the watershed. Time-aggregated data from five CTD (conductivity-temperature-depth) sensors in selected streams are used for calibration. The model outputs and the sensor data demonstrate a strong linear correlation, with the R value of 0.97. The investigation shows that the static modeling approach may provide an effective method for the understanding the input and transport of road salt to within watersheds.

  19. The HAW-Project: Test disposal of highly radioactive radiation sources in the Asse salt mine

    International Nuclear Information System (INIS)

    1992-04-01

    Two electrical heater tests were already started in November 1988 and are continuously surveyed in respect of the thermomechanical and geochemical response of the rock mass. Also the handling system necessary for the emplacement of 30 radioactive canisters (Sr-90 and Cs-137 sources) was developed and succesfully tested. This system consists of six multiple transport and storage casks of the type Castor-GSF-5, two above ground/below ground shuttle transport casks of the type Asse TB1, an above ground transfer station, an underground transport vehicle, a disposal machine, and a borehole slider. A laboratory investigation program on radiation effects in salt is being performed in advance to the radioactive canister emplacement. This program includes the investigation of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. For gamma dose and dose rate measurements in the test field measuring systems consisting of ionisation chambers as well as solid state dosemeters were developed and tested. Thermomechanical computer code validation is performed by calculational predictions and parallel investigation of the stress and displacement fields in the underground test field. (orig./HP)

  20. Computer access security code system

    Science.gov (United States)

    Collins, Earl R., Jr. (Inventor)

    1990-01-01

    A security code system for controlling access to computer and computer-controlled entry situations comprises a plurality of subsets of alpha-numeric characters disposed in random order in matrices of at least two dimensions forming theoretical rectangles, cubes, etc., such that when access is desired, at least one pair of previously unused character subsets not found in the same row or column of the matrix is chosen at random and transmitted by the computer. The proper response to gain access is transmittal of subsets which complete the rectangle, and/or a parallelepiped whose opposite corners were defined by first groups of code. Once used, subsets are not used again to absolutely defeat unauthorized access by eavesdropping, and the like.

  1. Remote sensing and geographic information system for appraisal of salt-affected soils in India.

    Science.gov (United States)

    Singh, Gurbachan; Bundela, D S; Sethi, Madhurama; Lal, Khajanchi; Kamra, S K

    2010-01-01

    Quantification of the nature, extent, and spatial distribution of salt-affected soils (SAS) for India and the world is essential for planning and implementing reclamation programs in a timely and cost-effective manner for sustained crop production. The national extent of SAS for India over the last four decades was assessed by conventional and remote sensing approaches using diverse methodologies and class definitions and ranged from 6.0 to 26.1 million hectares (Mha) and 1.2 to 10.1 Mha, respectively. In 1966, an area of 6 Mha under SAS was first reported using the former approach. Three national estimates, obtained using remote sensing, were reconciled using a geographic information system, resulting in an acceptable extent of 6.73 Mha. Moderately and severely salt-encrusted lands having large contiguous area have been correctly mapped, but slightly salt-encrusted land having smaller affected areas within croplands has not been accurately mapped. Recent satellite sensors (e.g., Resourcesat-1, Cartosat-2, IKONOS-II, and RISAT-2), along with improved image processing techniques integrated with terrain and other spatial data using a geographic information system, are enabling mapping at large scale. Significant variations in salt encrustation at the surface caused by soil moisture, waterlogging conditions, salt-tolerant crops, and dynamics of subsurface salts present constraints in appraisal, delineation, and mapping efforts. The article provides an overview of development, identification, characterization, and delineation of SAS, past and current national scenarios of SAS using conventional and remote sensing approaches, reconciliation of national estimates, issues of SAS mapping, and future scope.

  2. Performance analysis of multiple interference suppression over asynchronous/synchronous optical code-division multiple-access system based on complementary/prime/shifted coding scheme

    Science.gov (United States)

    Nieh, Ta-Chun; Yang, Chao-Chin; Huang, Jen-Fa

    2011-08-01

    A complete complementary/prime/shifted prime (CPS) code family for the optical code-division multiple-access (OCDMA) system is proposed. Based on the ability of complete complementary (CC) code, the multiple-access interference (MAI) can be suppressed and eliminated via spectral amplitude coding (SAC) OCDMA system under asynchronous/synchronous transmission. By utilizing the shifted prime (SP) code in the SAC scheme, the hardware implementation of encoder/decoder can be simplified with a reduced number of optical components, such as arrayed waveguide grating (AWG) and fiber Bragg grating (FBG). This system has a superior performance as compared to previous bipolar-bipolar coding OCDMA systems.

  3. Validation of the VTT's reactor physics code system

    International Nuclear Information System (INIS)

    Tanskanen, A.

    1998-01-01

    At VTT Energy several international reactor physics codes and nuclear data libraries are used in a variety of applications. The codes and libraries are under constant development and every now and then new updated versions are released, which are taken in use as soon as they have been validated at VTT Energy. The primary aim of the validation is to ensure that the code works properly, and that it can be used correctly. Moreover, the applicability of the codes and libraries are studied in order to establish their advantages and weak points. The capability of generating program-specific nuclear data for different reactor physics codes starting from the same evaluated data is sometimes of great benefit. VTT Energy has acquired a nuclear data processing system based on the NJOY-94.105 and TRANSX-2.15 processing codes. The validity of the processing system has been demonstrated by generating pointwise (MCNP) and groupwise (ANISN) temperature-dependent cross section sets for the benchmark calculations of the Doppler coefficient of reactivity. At VTT Energy the KENO-VI three-dimensional Monte Carlo code is used in criticality safety analyses. The KENO-VI code and the 44GROUPNDF5 data library have been validated at VTT Energy against the ZR-6 and LR-0 critical experiments. Burnup Credit refers to the reduction in reactivity of burned nuclear fuel due to the change in composition during irradiation. VTT Energy has participated in the calculational VVER-440 burnup credit benchmark in order to validate criticality safety calculation tools. (orig.)

  4. V.S.O.P. (99/05) computer code system

    International Nuclear Information System (INIS)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Scherer, W.

    2005-11-01

    V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99 / 05) represents the further development of V.S.O.P. (99). Compared to its precursor, the code system has been improved in many details. Major improvements and extensions have been included concerning the neutron spectrum calculation, the 3-d neutron diffusion options, and the thermal hydraulic section with respect to 'multi-pass'-fuelled pebblebed cores. This latest code version was developed and tested under the WINDOWS-XP - operating system. The storage requirement for the executables and the basic libraries associated with the code amounts to about 15 MB. Another 5 MB are required - if desired - for storage of the source code (∼65000 Fortran statements). (orig.)

  5. V.S.O.P. (99/05) computer code system

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Scherer, W.

    2005-11-01

    V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99 / 05) represents the further development of V.S.O.P. (99). Compared to its precursor, the code system has been improved in many details. Major improvements and extensions have been included concerning the neutron spectrum calculation, the 3-d neutron diffusion options, and the thermal hydraulic section with respect to 'multi-pass'-fuelled pebblebed cores. This latest code version was developed and tested under the WINDOWS-XP - operating system. The storage requirement for the executables and the basic libraries associated with the code amounts to about 15 MB. Another 5 MB are required - if desired - for storage of the source code ({approx}65000 Fortran statements). (orig.)

  6. Development of FBR integrity system code. Basic concept

    International Nuclear Information System (INIS)

    Asayama, Tai

    2001-05-01

    For fast breeder reactors to be commercialized, they must be more reliable, safer, and at the same, economically competitive with future light water reactors. Innovation of elevated temperature structural design standard is necessary to achieve this goal. The most powerful way is to enlarge the scope of structural integrity code to cover items other than design evaluation that has been addressed in existing codes. Items that must be newly covered are prerequisites of design, fabrication, examination, operation and maintenance, etc. This allows designers to choose the most economical combination of design variations to achieve specific reliability that is needed for a particular component. Designing components by this concept, a cost-minimum design of a whole plant can be realized. By determining the reliability that must be achieved for a component by risk technologies, further economical improvement can be expected by avoiding excessive quality. Recognizing the necessity for the codes based on the new concept, the development of 'FBR integrity system code' began in 2000. Research and development will last 10 years. For this development, the basic logistics and system as well as technologies that materialize the concept are necessary. Original logistics and system must be developed, because no existing researches are available in and out of Japan. This reports presents the results of the work done in the first year regarding the basic idea, methodology, and structure of the code. (author)

  7. 14 CFR Sec. 1-4 - System of accounts coding.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false System of accounts coding. Sec. 1-4 Section... General Accounting Provisions Sec. 1-4 System of accounts coding. (a) A four digit control number is assigned for each balance sheet and profit and loss account. Each balance sheet account is numbered...

  8. SWAT2: The improved SWAT code system by incorporating the continuous energy Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Mochizuki, Hiroki; Suyama, Kenya; Okuno, Hiroshi

    2003-01-01

    SWAT is a code system, which performs the burnup calculation by the combination of the neutronics calculation code, SRAC95 and the one group burnup calculation code, ORIGEN2.1. The SWAT code system can deal with the cell geometry in SRAC95. However, a precise treatment of resonance absorptions by the SRAC95 code using the ultra-fine group cross section library is not directly applicable to two- or three-dimensional geometry models, because of restrictions in SRAC95. To overcome this problem, SWAT2 which newly introduced the continuous energy Monte Carlo code, MVP into SWAT was developed. Thereby, the burnup calculation by the continuous energy in any geometry became possible. Moreover, using the 147 group cross section library called SWAT library, the reactions which are not dealt with by SRAC95 and MVP can be treated. OECD/NEA burnup credit criticality safety benchmark problems Phase-IB (PWR, a single pin cell model) and Phase-IIIB (BWR, fuel assembly model) were calculated as a verification of SWAT2, and the results were compared with the average values of calculation results of burnup calculation code of each organization. Through two benchmark problems, it was confirmed that SWAT2 was applicable to the burnup calculation of the complicated geometry. (author)

  9. HTR core physics and transient analyses by the Panthermix code system

    Energy Technology Data Exchange (ETDEWEB)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

  10. Particle and heavy ion transport code system; PHITS

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

    Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear

  11. Extension of the EQ3/6 computer codes to geochemical modeling of brines

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, K.J.; Wolery, T.J.

    1984-10-23

    Recent modifications to the EQ3/6 geochemical modeling software package provide for the use of Pitzer's equations to calculate the activity coefficients of aqueous species and the activity of water. These changes extend the range of solute concentrations over which the codes can be used to dependably calculate equilibria in geochemical systems, and permit the inclusion of ion pairs, complexes, and undissociated acids and bases as explicit component species in the Pitzer model. Comparisons of calculations made by the EQ3NR and EQ6 compuer codes with experimental data confirm that the modifications not only allow the codes to accurately evaluate activity coefficients in concentrated solutions, but also permit prediction of solubility limits of evaporite minerals in brines at 25/sup 0/C and elevated temperatures. Calculations for a few salts can be made at temperatures up to approx. 300/sup 0/C, but the temperature range for most electrolytes is constrained by the availability of requisite data to values less than or equal to 100/sup 0/C. The implementation of Pitzer's equations in EQ3/6 allows application of these codes to problems involving calculation of geochemical equilibria in brines; such as evaluation of the chemical environment which might be anticipated for nuclear waste canisters located in a salt repository. 26 references, 3 figures, 1 table.

  12. Development of High Temperature Transport System for Molten Salt

    International Nuclear Information System (INIS)

    Lee, S. H.; Lee, H. S.; Kim, J. G.

    2011-01-01

    Pyroprocessing technology is one of the the most promising technologies for the advanced fuel cycle with favorable economic potential and intrinsic proliferation-resistance. The electrorefining process, one of main processes which is composed of pyroprocess to recover the useful elements from spent fuel, is under development at the Korea Atomic Energy Research Institute as a sub process of the pyrochemical treatment of spent PWR fuel. High-temperature molten salt transport technologies are required because a molten salt should be transported from the electrorefiner to electrowiner after the electrorefining process. Therefore, in pyrometallurgical processing, the development of high-temperature molten salt transport technologies is a crucial prerequisite. However, there have been a few transport studies on high-temperature molten salt. In this study, an apparatus for suction transport experiments was designed and constructed for the development of high temperature transport technology for molten salt, and the performance test of the apparatus was performed. And also, predissolution test of the salt was carried out using the reactor with furnace in experimental apparatus

  13. Los Alamos radiation transport code system on desktop computing platforms

    International Nuclear Information System (INIS)

    Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.; West, J.T.

    1990-01-01

    The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. The current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines

  14. RELAP5-3D Code Includes ATHENA Features and Models

    International Nuclear Information System (INIS)

    Riemke, Richard A.; Davis, Cliff B.; Schultz, Richard R.

    2006-01-01

    Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, carbon dioxide, carbon monoxide, helium, hydrogen, krypton, nitrogen, oxygen, SF 6 , xenon) can be specified as part of the vapor/gas phase of the working fluid. These noncondensable gases were in previous versions of RELAP5-3D. Recently four molten salts have been added as working fluids to RELAP5-3D Version 2.4, which has had limited release. These molten salts will be in RELAP5-3D Version 2.5, which will have a general release like RELAP5-3D Version 2.3. Applications that use these new features and models are discussed in this paper. (authors)

  15. Systems costs for disposal of Savannah River high-level waste sludge and salt

    International Nuclear Information System (INIS)

    McDonell, W.R.; Goodlett, C.B.

    1984-01-01

    A systems cost model has been developed to support disposal of defense high-level waste sludge and salt generated at the Savannah River Plant. Waste processing activities covered by the model include decontamination of the salt by a precipitation process in the waste storage tanks, incorporation of the sludge and radionuclides removed from the salt into glass in the Defense Waste Processing Facility (DWPF), and, after interim storage, final disposal of the DWPF glass waste canisters in a federal geologic repository. Total costs for processing of waste generated to the year 2000 are estimated to be about $2.9 billion (1984 dollars); incremental unit costs for DWPF and repository disposal activities range from $120,000 to $170,000 per canister depending on DWPF processing schedules. In a representative evaluation of process alternatives, the model is used to demonstrate cost effectiveness of adjustments in the frit content of the waste glass to reduce impacts of wastes generated by the salt decontamination operations. 13 references, 8 tables

  16. Salt Effect on the Cloud Point Phenomenon of Amphiphilic Drug-Hydroxypropylmethyl Cellulose System

    Directory of Open Access Journals (Sweden)

    Mohd. Sajid Ali

    2014-01-01

    Full Text Available Effect of two amphiphilic drugs (tricyclic antidepressant, nortriptyline hydrochloride (NORT, and nonsteroidal anti-inflammatory drug, sodium salt of ibuprofen (IBF on the cloud point of biopolymer hydroxypropylmethyl cellulose (HPMC was studied. Effect of NaCl was also seen on the CP of HPMC-drug system. CP of HPMC increases uniformly on increasing the (drug. Both drugs, though one being anionic (IBF and other cationic (NORT, affect the CP in almost the same manner but with different extent implying the role of hydrophobicity in the interaction between drug and polymer. Salt affects the CP of the drug in a dramatic way as low concentration of salt was only able to increase the value of the CP, though not affecting the pattern. However, in presence of high concentration of salts, minimum was observed on CP versus (drug plots. Various thermodynamic parameters were evaluated and discussed on the basis of the observed results.

  17. Application of startup/core management code system to YGN 3 startup testing

    International Nuclear Information System (INIS)

    Chi, Sung Goo; Hah, Yung Joon; Doo, Jin Yong; Kim, Dae Kyum

    1995-01-01

    YGN 3 is the first nuclear power plant in Korea to use the fixed incore detector system for startup testing and core management. The startup/core management code system was developed from existing ABB-C-E codes and applied for YGN 3 startup testing, especially for physics and CPC(Core Protection Calculator)/COLSS (Core Operating Limit Supervisory System) related testing. The startup/core management code system consists of startup codes which include the CEBASE, CECOR, CEFAST and CEDOPS, and startup data reduction codes which include FLOWRATE, COREPERF, CALMET, and VARTAV. These codes were implemented on an HP/Apollo model 9000 series 400 workstation at the YGN 3 site and successfully applied to startup testing and core management. The startup codes made a great contribution in upgrading the reliability of test results and reducing the test period by taking and analyzing core data automatically. The data reduction code saved the manpower and time for test data reduction and decreased the chance for error in the analysis. It is expected that this code system will make similar contributions for reducing the startup testing duration of YGN 4 and UCN3,4

  18. Systems engineering management plan for the Salt Repository Project

    International Nuclear Information System (INIS)

    Neff, J.O.

    1986-08-01

    This document presents the plan for using systems engineering in conducting and managing the technical work of the Salt Repository Project (SRP) of the US Department of Energy's Civilian Radioactive Waste Management Program. The need for preparing a Systems Engineering Management Plan (SEMP) is traced back to relevant DOE directives. These directives are interpreted as SRP requirements in the context of the Mined Geologic Disposal System. The strategy for conducting systems engineering on the SRP, including the role of the systems engineering process, is then described. The SEMP also designates who in the project organization will be responsible for carrying out the activities. Finally, the management tools that are used to implement the systems engineering process, including associated documentation on the SRP, are described

  19. Integrated demonstration of molten salt oxidation with salt recycle for mixed waste treatment

    International Nuclear Information System (INIS)

    Hsu, P.C.

    1997-01-01

    Molten Salt Oxidation (MSO) is a thermal, nonflame process that has the inherent capability of completely destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility and constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are performed under carefully controlled (experimental) conditions. The system consists of a MSO processor with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. This integrated system was designed and engineered based on laboratory experience with a smaller engineering-scale reactor unit and extensive laboratory development on salt recycle and final forms preparation. In this paper we present design and engineering details of the system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is identification of the most suitable waste streams and waste types for MSO treatment

  20. [Data coding in the Israeli healthcare system - do choices provide the answers to our system's needs?].

    Science.gov (United States)

    Zelingher, Julian; Ash, Nachman

    2013-05-01

    The IsraeLi healthcare system has undergone major processes for the adoption of health information technologies (HIT), and enjoys high Levels of utilization in hospital and ambulatory care. Coding is an essential infrastructure component of HIT, and ts purpose is to represent data in a simplified and common format, enhancing its manipulation by digital systems. Proper coding of data enables efficient identification, storage, retrieval and communication of data. UtiLization of uniform coding systems by different organizations enables data interoperability between them, facilitating communication and integrating data elements originating in different information systems from various organizations. Current needs in Israel for heaLth data coding include recording and reporting of diagnoses for hospitalized patients, outpatients and visitors of the Emergency Department, coding of procedures and operations, coding of pathology findings, reporting of discharge diagnoses and causes of death, billing codes, organizational data warehouses and national registries. New national projects for cLinicaL data integration, obligatory reporting of quality indicators and new Ministry of Health (MOH) requirements for HIT necessitate a high Level of interoperability that can be achieved only through the adoption of uniform coding. Additional pressures were introduced by the USA decision to stop the maintenance of the ICD-9-CM codes that are also used by Israeli healthcare, and the adoption of ICD-10-C and ICD-10-PCS as the main coding system for billing purpose. The USA has also mandated utilization of SNOMED-CT as the coding terminology for the ELectronic Health Record problem list, and for reporting quality indicators to the CMS. Hence, the Israeli MOH has recently decided that discharge diagnoses will be reported using ICD-10-CM codes, and SNOMED-CT will be used to code the cLinical information in the EHR. We reviewed the characteristics, strengths and weaknesses of these two coding

  1. Geographic Information Systems using CODES linked data (Crash outcome data evaluation system)

    Science.gov (United States)

    2001-04-01

    This report presents information about geographic information systems (GIS) and CODES linked data. Section one provides an overview of a GIS and the benefits of linking to CODES. Section two outlines the basic issues relative to the types of map data...

  2. Sub-channel/system coupled code development and its application to SCWR-FQT loop

    International Nuclear Information System (INIS)

    Liu, X.J.; Cheng, X.

    2015-01-01

    Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal–hydraulic parameters are predicted by the sub-channel code COBRA-SC. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel bundle can be reduced effectively by the safety measures

  3. Sub-channel/system coupled code development and its application to SCWR-FQT loop

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany)

    2015-04-15

    Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal–hydraulic parameters are predicted by the sub-channel code COBRA-SC. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel bundle can be reduced effectively by the safety measures

  4. Development of the vacuum system pressure responce analysis code PRAC

    International Nuclear Information System (INIS)

    Horie, Tomoyoshi; Kawasaki, Kouzou; Noshiroya, Shyoji; Koizumi, Jun-ichi.

    1985-03-01

    In this report, we show the method and numerical results of the vacuum system pressure responce analysis code. Since fusion apparatus is made up of many vacuum components, it is required to analyze pressure responce at any points of the system when vacuum system is designed or evaluated. For that purpose evaluating by theoretical solution is insufficient. Numerical analysis procedure such as finite difference method is usefull. In the PRAC code (Pressure Responce Analysis Code), pressure responce is obtained solving derivative equations which is obtained from the equilibrium relation of throughputs and contain the time derivative of pressure. As it considers both molecular and viscous flows, the coefficients of the equation depend on the pressure and the equations become non-linear. This non-linearity is treated as piece-wise linear within each time step. Verification of the code is performed for the simple problems. The agreement between numerical and theoretical solutions is good. To compare with the measured results, complicated model of gas puffing system is analyzed. The agreement is well for practical use. This code will be a useful analytical tool for designing and evaluating vacuum systems such as fusion apparatus. (author)

  5. Molten Salt Power Tower Cost Model for the System Advisor Model (SAM)

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, C. S.; Heath, G. A.

    2013-02-01

    This report describes a component-based cost model developed for molten-salt power tower solar power plants. The cost model was developed by the National Renewable Energy Laboratory (NREL), using data from several prior studies, including a contracted analysis from WorleyParsons Group, which is included herein as an Appendix. The WorleyParsons' analysis also estimated material composition and mass for the plant to facilitate a life cycle analysis of the molten salt power tower technology. Details of the life cycle assessment have been published elsewhere. The cost model provides a reference plant that interfaces with NREL's System Advisor Model or SAM. The reference plant assumes a nominal 100-MWe (net) power tower running with a nitrate salt heat transfer fluid (HTF). Thermal energy storage is provided by direct storage of the HTF in a two-tank system. The design assumes dry-cooling. The model includes a spreadsheet that interfaces with SAM via the Excel Exchange option in SAM. The spreadsheet allows users to estimate the costs of different-size plants and to take into account changes in commodity prices. This report and the accompanying Excel spreadsheet can be downloaded at https://sam.nrel.gov/cost.

  6. Evaluation and implementation of QR Code Identity Tag system for Healthcare in Turkey

    OpenAIRE

    Uzun, Vassilya; Bilgin, Sami

    2016-01-01

    For this study, we designed a QR Code Identity Tag system to integrate into the Turkish healthcare system. This system provides QR code-based medical identification alerts and an in-hospital patient identification system. Every member of the medical system is assigned a unique QR Code Tag; to facilitate medical identification alerts, the QR Code Identity Tag can be worn as a bracelet or necklace or carried as an ID card. Patients must always possess the QR Code Identity bracelets within hospi...

  7. Spanish participation in the Haw Project: Laboratory investigations on Gamma irradiation effects in rock salt

    International Nuclear Information System (INIS)

    Cuevas, C. de las; Miralles, L.; Teixidor, P.; Garcia Veigas, J.; Dies, X.; Ortega, X.; Pueyo, J.J.

    1993-01-01

    In order to prove the safe disposal of high-level radioactive waste (HAW) in salt rock, a five years test disposal of thirty highly radioactive radiation sources is planned in the Asse salt mine, in the Federal Republic of Germany. The thirty radiation sources consist of steel canisters containing the vitrified radionuclides Caesium 137 and Strontium 90 in quantities sufficient to cover the bandwidth of heat generation and gamma radiation of real HAW. The radiation sources will be emplaced in six boreholes located in two galleries at the 800 m level. Two electrical heater tests were already started in November 1988 and are continuosly surveyed in respect of the rock mass. Also the handling system necessary for the emplacement of the radioactive canisters was developed and succesfully tested. A laboratory investigation programme on radiation effects in salt is being performed in advance to the radioactive canister emplacement. This programme includes the investigation of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Part of this programme has been carried out since 1988 at the University of Barcelona, basically what refers to colloidal sodium determinations by light absorption measurements and microstructural studies on irradiated salt samples. For gamma dose and dose rate measurements in the test field, measuring systems consisting of ionisation chambers as well as solid state dosemeters were developed and tested. Thermomechanical computer code validation is performed by calculational predictions and parallel investigation of the stress and displacement fields in the underground test field

  8. Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

    Directory of Open Access Journals (Sweden)

    Alessandro Petruzzi

    2008-01-01

    Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.

  9. SRAC95; general purpose neutronics code system

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Tsuchihashi, Keichiro; Kaneko, Kunio.

    1996-03-01

    SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author)

  10. SRAC95; general purpose neutronics code system

    Energy Technology Data Exchange (ETDEWEB)

    Okumura, Keisuke; Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-03-01

    SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author).

  11. Solvation-based vapour pressure model for (solvent + salt) systems in conjunction with the Antoine equation

    International Nuclear Information System (INIS)

    Senol, Aynur

    2013-01-01

    Highlights: • Vapour pressures of (solvent + salt) systems have been estimated through a solvation-based model. • Two structural forms of the generalized solvation model using the Antoine equation have been performed. • A simplified concentration-dependent vapour pressure model has been also processed. • The model reliability analysis has been performed in terms of a log-ratio objective function. • The reliability of the models has been interpreted in terms of the statistical design factors. -- Abstract: This study deals with modelling the vapour pressure of a (solvent + salt) system on the basis of the principles of LSER. The solvation model framework clarifies the simultaneous impact of several physical variables such as the vapour pressure of a pure solvent estimated by the Antoine equation, the solubility and solvatochromic parameters of the solvent and the physical properties of the ionic salt. It has been analyzed independently the performance of two structural forms of the generalized model, i.e., a relation depending on an integration of the properties of the solvent and the ionic salt and a relation on a reduced property-basis. A simplified concentration-dependent vapour pressure model has been also explored and implemented on the relevant systems. The vapour pressure data of sixteen (solvent + salt) systems have been processed to analyze statistically the reliability of existing models in terms of a log–ratio objective function. The proposed vapour pressure models match relatively well the observed performance, yielding the overall design factors of 1.066 and 1.073 for the solvation-based models with the integrated and reduced properties, and 1.008 for the concentration-based model, respectively

  12. Development of the integrated system reliability analysis code MODULE

    International Nuclear Information System (INIS)

    Han, S.H.; Yoo, K.J.; Kim, T.W.

    1987-01-01

    The major components in a system reliability analysis are the determination of cut sets, importance measure, and uncertainty analysis. Various computer codes have been used for these purposes. For example, SETS and FTAP are used to determine cut sets; Importance for importance calculations; and Sample, CONINT, and MOCUP for uncertainty analysis. There have been problems when the codes run each other and the input and output are not linked, which could result in errors when preparing input for each code. The code MODULE was developed to carry out the above calculations simultaneously without linking input and outputs to other codes. MODULE can also prepare input for SETS for the case of a large fault tree that cannot be handled by MODULE. The flow diagram of the MODULE code is shown. To verify the MODULE code, two examples are selected and the results and computation times are compared with those of SETS, FTAP, CONINT, and MOCUP on both Cyber 170-875 and IBM PC/AT. Two examples are fault trees of the auxiliary feedwater system (AFWS) of Korea Nuclear Units (KNU)-1 and -2, which have 54 gates and 115 events, 39 gates and 92 events, respectively. The MODULE code has the advantage that it can calculate the cut sets, importances, and uncertainties in a single run with little increase in computing time over other codes and that it can be used in personal computers

  13. Quality assurance and verification of the MACCS [MELCOR Accident Consequence Code System] code, Version 1.5

    International Nuclear Information System (INIS)

    Dobbe, C.A.; Carlson, E.R.; Marshall, N.H.; Marwil, E.S.; Tolli, J.E.

    1990-02-01

    An independent quality assurance (QA) and verification of Version 1.5 of the MELCOR Accident Consequence Code System (MACCS) was performed. The QA and verification involved examination of the code and associated documentation for consistent and correct implementation of the models in an error-free FORTRAN computer code. The QA and verification was not intended to determine either the adequacy or appropriateness of the models that are used MACCS 1.5. The reviews uncovered errors which were fixed by the SNL MACCS code development staff prior to the release of MACCS 1.5. Some difficulties related to documentation improvement and code restructuring are also presented. The QA and verification process concluded that Version 1.5 of the MACCS code, within the scope and limitations process concluded that Version 1.5 of the MACCS code, within the scope and limitations of the models implemented in the code is essentially error free and ready for widespread use. 15 refs., 11 tabs

  14. Coupled neutronics/thermal-hydraulics for analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Zhou, Jianjun; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

    Highlights: ► A multiple-channel analysis code (MAC) is developed to be coupled with MCNP. ► 1/8 of core is simulated in MCNP and thermal-hydraulic code. ► The coupling calculation can achieve stable state after a few iterations. ► The coupling calculation results are in reasonable agreement with the analytic solutions of the ORNL. ► Parametric studies of MSR are performed to provide valuable information for future design MSR. -- Abstract: The Generation IV International Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of molten salt reactor experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. For further design analysis, parametric studies are performed to provide valuable information for new design of MSR. The effect of inlet temperature, graphite to molten salt volume ratio (G/Ms) from varying channel diameter and different power levels on the effective multiplication factor, neutron flux, graphite lifetime and temperature distribution are discussed in detail

  15. Coupled CFD - system-code simulation of a gas cooled reactor

    International Nuclear Information System (INIS)

    Yan, Yizhou; Rizwan-uddin

    2011-01-01

    A generic coupled CFD - system-code thermal hydraulic simulation approach was developed based on FLUENT and RELAP-3D, and applied to LWRs. The flexibility of the coupling methodology enables its application to advanced nuclear energy systems. Gas Turbine - Modular Helium Reactor (GT-MHR) is a Gen IV reactor design which can benefit from this innovative coupled simulation approach. Mixing in the lower plenum of the GT-MHR is investigated here using the CFD - system-code coupled simulation tool. Results of coupled simulations are presented and discussed. The potential of the coupled CFD - system-code approach for next generation of nuclear power plants is demonstrated. (author)

  16. Neutron shielding studies on an advanced molten salt fast reactor design

    International Nuclear Information System (INIS)

    Merk, Bruno; Konheiser, Jörg

    2014-01-01

    Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem

  17. Aqueous-salt system containing ytterbium nitrate and pyridine nitrate

    International Nuclear Information System (INIS)

    Zhuravlev, E.F.; Khisaeva, D.A.; Izmajlova, L.V.

    1983-01-01

    Cross-section method has been used to study solubility in ternary aqueous-salt system Yb(NO 3 ) 3 -C 5 H 5 NxHNO 3 -H 2 0 at 25 and 50 deg C. It is established that the system is characterized by chemical interaction. Congruently soluble compound of Yb(NO 3 ) 3 x2[C 5 H 5 NxHNO 3 ] composition is discovered in the system. Composition of the compound is confirmed by chemical analysis; its infrared spectra are studied. Interplanar distances are determined; derivatogram of the compound is given. The results of the works are compared with analogous investigations of another rare earth nitrates

  18. Mined salt storage feasibility: Engineering study report

    International Nuclear Information System (INIS)

    1987-07-01

    This study addresses a method of eliminating the surface storage of mined salt at the Deaf Smith repository site. It provides rough estimates of the logistics and costs of transporting 3.7 million tons of salt from the repository to the salt disposal site near Carlsbad, New Mexico and returning it to the repository for decommissioning backfill. The study assumes that a railcar/truck system will be installed and that the excavated salt will be transported from the repository to an existing potash mine located near Carlsbad, New Mexico approximately 300 miles from the repository. The 3.7 million tons of salt required for repository decommissioning backfill can be stored in the potash mines along with the excess salt, with no additional capital costs required for either a railcar or a truck transportation system. The capital cost for facilities to reclaim the 3.7 million tons of salt from the potash mine is estimated to be $4,400,000 with either a rail or truck transportation system. Segregating the 3.7 million tons of backfill salt in a surface storage area at the potash mine requires a capital cost of $13,900,000 with a rail system or $11,400,000 with a truck system. Transportation costs are estimated at $0.08/ton-mile for rail and $0.13/ton-mile for truck. 2 figs., 5 tabs

  19. Differential Space-Time Block Code Modulation for DS-CDMA Systems

    Directory of Open Access Journals (Sweden)

    Liu Jianhua

    2002-01-01

    Full Text Available A differential space-time block code (DSTBC modulation scheme is used to improve the performance of DS-CDMA systems in fast time-dispersive fading channels. The resulting scheme is referred to as the differential space-time block code modulation for DS-CDMA (DSTBC-CDMA systems. The new modulation and demodulation schemes are especially studied for the down-link transmission of DS-CDMA systems. We present three demodulation schemes, referred to as the differential space-time block code Rake (D-Rake receiver, differential space-time block code deterministic (D-Det receiver, and differential space-time block code deterministic de-prefix (D-Det-DP receiver, respectively. The D-Det receiver exploits the known information of the spreading sequences and their delayed paths deterministically besides the Rake type combination; consequently, it can outperform the D-Rake receiver, which employs the Rake type combination only. The D-Det-DP receiver avoids the effect of intersymbol interference and hence can offer better performance than the D-Det receiver.

  20. A study on the nuclear computer codes installation and management system

    International Nuclear Information System (INIS)

    Kim, Yeon Seung; Huh, Young Hwan; Kim, Hee Kyung; Kang, Byung Heon; Kim, Ko Ryeo; Suh, Soong Hyok; Choi, Young Gil; Lee, Jong Bok

    1990-12-01

    From 1987 a number of technical transfer related to nuclear power plant had been performed from C-E for YGN 3 and 4 construction. Among them, installation and management of the computer codes for YGN 3 and 4 fuel and nuclear steam supply system was one of the most important project. Main objectives of this project are to establish the nuclear computer code management system, to develop QA procedure for nuclear codes, to secure the nuclear code reliability and to extend techanical applicabilities including the user-oriented utility programs for nuclear codes. Contents of performing the project in this year was to produce 215 transmittal packages of nuclear codes installation including making backup magnetic tape and microfiche for software quality assurance. Lastly, for easy reference about the nuclear codes information we presented list of code names and information on the codes which were introduced from C-E. (Author)

  1. Fundamental study on the salt distillation from the mixtures of rare earth precipitates and LiCl-KCl eutectic salt

    International Nuclear Information System (INIS)

    Yang, H. C.; Eun, H. C.; Cho, Y. Z.; Lee, H. S.; Kim, I. T.

    2008-01-01

    An electrorefining process of spent nuclear fuel generates waste salt containing some radioactive metal chlorides. The most effective method to reduce salt waste volume is to separate radioactive metals from non-radioactive salts. A promising approach is to change radioactive metal chlorides into salt-insoluble oxides by an oxygen sparging. Following this, salt distillation process is available to effectively separate the precipitated particulate metal oxides from salt. This study investigated the distillation rates of LiCl-KCl eutectic salt under different vacuums at elevated temperatures. The first part study investigated distillation rates of eutectic salt under different vacuums at high temperatures by using thermo-gravimetric furnace system. In the second part, we tested the removal of eutectic salt from the RE precipitates by using the laboratory vacuum distillation furnace system. Investigated variables were the temperature of mixture, the degree of vacuum and the time

  2. System verification and validation report for the TMAD code

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    1995-01-01

    This document serves as the Verification and Validation Report for the TMAD code system, which includes the TMAD code and the LIBMAKR Code. The TMAD code was commissioned to facilitate the interpretation of moisture probe measurements in the Hanford Site waste tanks. In principle, the code is an interpolation routine that acts over a library of benchmark data based on two independent variables, typically anomaly size and moisture content. Two additional variables, anomaly type and detector type, can also be considered independent variables, but no interpolation is done over them. The dependent variable is detector response. The intent is to provide the code with measured detector responses from two or more detectors. The code will then interrogate (and interpolate upon) the benchmark data library and find the anomaly-type/anomaly-size/moisture-content combination that provides the closest match to the measured data. The primary purpose of this document is to provide the results of the system testing and the conclusions based thereon. The results of the testing process are documented in the body of the report. Appendix A gives the test plan, including test procedures, used in conducting the tests. Appendix B lists the input data required to conduct the tests, and Appendices C and 0 list the numerical results of the tests

  3. Proposal on the accelerator driven molten-salt reactor (ATW-concept) benchmark calculation (stage-1 without an external neutron sources)

    International Nuclear Information System (INIS)

    Svarny, J.; Mikolas, P.

    1999-01-01

    The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark will be not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (Authors)

  4. A way to limit the corrosion in the Molten Salt Reactor concept: the salt redox potential control

    International Nuclear Information System (INIS)

    Gibilaro, M.; Massot, L.; Chamelot, P.

    2015-01-01

    The possibility of controlling the salt redox potential thanks to a redox buffer in the Molten Salt Fast Reactor was investigated, the goal was to limit the oxidation of the reactor structural material. Tests were performed in LiF-CaF 2 at 850 °C on two different redox couples to fix the salt potential, Eu(III)/Eu(II) and U(IV)/U(III), where the first one was used as inactive system to validate the methodology to be applied on the uranium system. A metallic reducing agent (Gd plate for Eu, and U plate for U system) was inserted in the salt, leading to a spontaneous reaction: Eu(III) and U(IV) were then reduced. Eu(III) was fully converted into Eu(II) with metallic Gd, validating the approach. On the U system, the U(IV)/U(III) ratio has to be set between 10 and 100 to limit the core material oxidation: addition of metallic U decreased the concentration ratio from the infinite to 1, showing the feasibility of the salt redox potential control with the U system

  5. Perspectives on the development of next generation reactor systems safety analysis codes

    International Nuclear Information System (INIS)

    Zhang, H.

    2015-01-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  6. Perspectives on the development of next generation reactor systems safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  7. Simulation of Molten Salt Reactor dynamics

    International Nuclear Information System (INIS)

    Krepel, J.; Rohde, U.; Grundmann, U.

    2005-01-01

    Dynamics of the Molten Salt Reactor - one of the 'Generation IV' concepts - was studied in this paper. The graphite-moderated channel type MSR was selected for the numerical simulation of the reactor with liquid fuel. The MSR dynamics is very specific because of two physical peculiarities of the liquid fueled reactor: the delayed neutrons precursors are drifted by the fuel flow and the fission energy is immediately released directly into the coolant. Presently, there are not many accessible numerical codes appropriate for the MSR simulation, therefore the DYN3D-MSR code was developed based on the FZR in-house code DYN3D. It allows calculating of full 3D transient neutronics in combination with parallel channel type thermal-hydraulics. By means of DYN3D-MSR, several transients typical for the liquid fuel system were analyzed. Those transients were initiated by reactivity insertion, by overcooling of fuel at the core inlet, by the fuel pump start-up or coast-down, or by the blockage of selected fuel channels. In these considered transients, the response of the MSR is characterized by the immediate change of the fuel temperature with changing power and fast negative temperature feedback to the power. The response through the graphite temperature is slower. Furthermore, for big MSR cores fueled with U233 the graphite feedback coefficient can be positive. In this case the addition of erbium to the graphite can ensure the inherent safety features. The DYN3D-MSR code has been shown to be an effective tool for MSR dynamics studies. (author)

  8. The FRAM code: Description and some comparisons with MGA

    International Nuclear Information System (INIS)

    Sampson, T.E.; Kelley, T.A.

    1994-01-01

    The authors describe the initial development of the FRAM gamma-ray spectrometry code for analyzing plutonium isotopics, discuss its methodology, and present some comparisons with MGA on identical items. They also present some of the features of a new Windows 3.1-based version (PC/FRAM) and describe some current measurement problems. Development of the FRAM code began in about 1985, growing out of the need at the Los Alamos TA-55 Plutonium Facility for an isotopic analysis code to give accurate results for the effective specific power of heterogeneous (Am/Pu) pyrochemical residues. These residues present a difficult challenge because the americium is present mostly in a low-Z salt matrix (AmCl 3 ) with fines and small pieces of plutonium metal dispersed throughout the salt. Plutonium gamma rays suffer different attenuation than americium gamma rays of the same energy; this makes conventional analysis with a single relative efficiency function inaccurate for Am/Pu ratios and affects the analysis in other subtle ways

  9. Coded aperture imaging system for nuclear fuel motion detection

    International Nuclear Information System (INIS)

    Stalker, K.T.; Kelly, J.G.

    1980-01-01

    A Coded Aperature Imaging System (CAIS) has been developed at Sandia National Laboratories to image the motion of nuclear fuel rods undergoing tests simulating accident conditions within a liquid metal fast breeder reactor. The tests require that the motion of the test fuel be monitored while it is immersed in a liquid sodium coolant precluding the use of normal optical means of imaging. However, using the fission gamma rays emitted by the fuel itself and coded aperture techniques, images with 1.5 mm radial and 5 mm axial resolution have been attained. Using an electro-optical detection system coupled to a high speed motion picture camera a time resolution of one millisecond can be achieved. This paper will discuss the application of coded aperture imaging to the problem, including the design of the one-dimensional Fresnel zone plate apertures used and the special problems arising from the reactor environment and use of high energy gamma ray photons to form the coded image. Also to be discussed will be the reconstruction techniques employed and the effect of various noise sources on system performance. Finally, some experimental results obtained using the system will be presented

  10. Balancing sub- and supra-salt strain in salt-influenced rifts: Implications for extension estimates

    Science.gov (United States)

    Coleman, Alexander J.; Jackson, Christopher A.-L.; Duffy, Oliver B.

    2017-09-01

    The structural style of salt-influenced rifts may differ from those formed in predominantly brittle crust. Salt can decouple sub- and supra-salt strain, causing sub-salt faults to be geometrically decoupled from, but kinematically coupled to and responsible for, supra-salt forced folding. Salt-influenced rifts thus contain more folds than their brittle counterparts, an observation often ignored in extension estimates. Fundamental to determining whether sub- and supra-salt structures are kinematically coherent, and the relative contributions of thin- (i.e. gravity-driven) and thick-skinned (i.e. whole-plate stretching) deformation to accommodating rift-related strain, is our ability to measure extension at both structural levels. We here use published physical models of salt-influenced extension to show that line-length estimates yield more accurate values of sub- and supra-salt extension compared to fault-heave, before applying these methods to seismic data from the Halten Terrace, offshore Norway. We show that, given the abundance of ductile deformation in salt-influenced rifts, significant amounts of extension may be ignored, leading to the erroneous interpretations of thin-skinned, gravity-gliding. If a system is kinematically coherent, supra-salt structures can help predict the occurrence and kinematics of sub-salt faults that may be poorly imaged and otherwise poorly constrained.

  11. Integrating bar-code devices with computerized MC and A systems

    International Nuclear Information System (INIS)

    Anderson, L.K.; Boor, M.G.; Hurford, J.M.

    1998-01-01

    Over the past seven years, Los Alamos National Laboratory developed several generations of computerized nuclear materials control and accountability (MC and A) systems for tracking and reporting the storage, movement, and management of nuclear materials at domestic and international facilities. During the same period, Oak Ridge National Laboratory was involved with automated data acquisition (ADA) equipment, including installation of numerous bar-code scanning stations at various facilities to serve as input devices to computerized systems. Bar-code readers, as well as other ADA devices, reduce input errors, provide faster input, and allow the capture of data in remote areas where workstations do not exist. Los Alamos National Laboratory and Oak Ridge National Laboratory teamed together to implement the integration of bar-code hardware technology with computerized MC and A systems. With the expertise of both sites, the two technologies were successfully merged with little difficulty. Bar-code input is now available with several functions of the MC and A systems: material movements within material balance areas (MBAs), material movements between MBAs, and physical inventory verification. This paper describes the various components required for the integration of these MC and A systems with the installed bar-code reader devices and the future directions for these technologies

  12. The in-core fuel management code system for VVER reactors

    International Nuclear Information System (INIS)

    Cada, R.; Krysl, V.; Mikolas, P.; Sustek, J.; Svarny, J.

    2004-01-01

    The structure and methodology of a fuel management system for NPP VVER 1000 (NPP Temelin) and VVER 440 (NPP Dukovany) is described. It is under development in SKODA JS a.s. and is followed by practical applications. The general objectives of the system are maximization of end of cycle reactivity, the minimization of fresh fuel inventory for the minimization of fed enrichment and minimization of burnable poisons (BPs) inventory. They are also safety related constraints in witch minimization of power peaking plays a dominant role. General structure of the system consists in preparation of input data for macrocode calculation, algorithms (codes) for optimization of fuel loading, calculation of fuel enrichment and BPs assignment. At present core loading can be calculated (optimized) by Tabu search algorithm (code ATHENA), genetic algorithm (code Gen1) and hybrid algorithm - simplex procedure with application of Tabu search algorithm on binary shuffling (code OPAL B ). Enrichment search is realized by the application of simplex algorithm (OPAL B code) and BPs assignment by module BPASS and simplex algorithm in OPAL B code. Calculations of the real core loadings are presented and a comparison of different optimization methods is provided. (author)

  13. Code system to compute radiation dose in human phantoms

    International Nuclear Information System (INIS)

    Ryman, J.C.; Cristy, M.; Eckerman, K.F.; Davis, J.L.; Tang, J.S.; Kerr, G.D.

    1986-01-01

    Monte Carlo photon transport code and a code using Monte Carlo integration of a point kernel have been revised to incorporate human phantom models for an adult female, juveniles of various ages, and a pregnant female at the end of the first trimester of pregnancy, in addition to the adult male used earlier. An analysis code has been developed for deriving recommended values of specific absorbed fractions of photon energy. The computer code system and calculational method are described, emphasizing recent improvements in methods

  14. Extended UNIQUAC Model for Correlation and Prediction of Vapor-Liquid-Liquid-Solid Equilibria in Aqueous Salt Systems Containing Non-Electrolytes. Part B. Alcohol (Ethanol, Propanols, Butanols) - Water-salt systems

    DEFF Research Database (Denmark)

    Thomsen, Kaj; Iliuta, Maria Cornelia; Rasmussen, Peter

    2004-01-01

    The Extended UNIQUAC model for electrolyte solutions is an excess Gibbs energy function consisting of a Debye-Huckel term and a term corresponding to the UNIQUAC equation. For vapor-liquid equilibrium calculations, the fugacities of gas-phase components are calculated with the Soave-Redlich-Kwong......The Extended UNIQUAC model for electrolyte solutions is an excess Gibbs energy function consisting of a Debye-Huckel term and a term corresponding to the UNIQUAC equation. For vapor-liquid equilibrium calculations, the fugacities of gas-phase components are calculated with the Soave...... solid-liquid-vapor equilibrium and thermal property data for strongly non-ideal systems. In this work, the model is extended to aqueous salt systems containing higher alcohols. The calculations are based on an extensive database consisting of salt solubility data, vapor liquid equilibrium data...

  15. LOLA SYSTEM: A code block for nodal PWR simulation. Part. II - MELON-3, CONCON and CONAXI Codes

    International Nuclear Information System (INIS)

    Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.

    1985-01-01

    Description of the theory and users manual of the MELON-3, CONCON and CONAXI codes, which are part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. These auxiliary codes, provide some of the input data for the main module SIMULA-3; these are, the reactivity correlations constants, the albe does and the transport factors. (Author) 7 refs

  16. Measurements of stimulated-Raman-scattering-induced tilt in spectral-amplitude-coding optical code-division multiple-access systems

    Science.gov (United States)

    Al-Qazwini, Zaineb A. T.; Abdullah, Mohamad K.; Mokhtar, Makhfudzah B.

    2009-01-01

    We measure the stimulated Raman scattering (SRS)-induced tilt in spectral-amplitude-coding optical code-division multiple-access (SAC-OCDMA) systems as a function of system main parameters (transmission distance, power per chip, and number of users) via computer simulations. The results show that SRS-induced tilt significantly increases as transmission distance, power per chip, or number of users grows.

  17. LOLA SYSTEM: A code block for nodal PWR simulation. Part. II - MELON-3, CONCON and CONAXI Codes

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J M; Ahnert, C; Gomez Santamaria, J; Rodriguez Olabarria, I

    1985-07-01

    Description of the theory and users manual of the MELON-3, CONCON and CONAXI codes, which are part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. These auxiliary codes, provide some of the input data for the main module SIMULA-3; these are, the reactivity correlations constants, the albe does and the transport factors. (Author) 7 refs.

  18. FAST: An advanced code system for fast reactor transient analysis

    International Nuclear Information System (INIS)

    Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh

    2005-01-01

    One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems

  19. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    International Nuclear Information System (INIS)

    Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

    2015-01-01

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF 2 -ThF 4 - 233 UF 4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155

  20. Modeling vapor pressures of solvent systems with and without a salt effect: An extension of the LSER approach

    International Nuclear Information System (INIS)

    Senol, Aynur

    2015-01-01

    Highlights: • A new polynomial vapor pressure approach for pure solvents is presented. • Solvation models reproduce the vapor pressure data within a 4% mean error. • A concentration-basis vapor pressure model is also implemented on relevant systems. • The reliability of existing models was analyzed using log-ratio objective function. - Abstract: A new polynomial vapor pressure approach for pure solvents is presented. The model is incorporated into the LSER (linear solvation energy relation) based solvation model framework and checked for consistency in reproducing experimental vapor pressures of salt-containing solvent systems. The developed two structural forms of the generalized solvation model (Senol, 2013) provide a relatively accurate description of the salting effect on vapor pressure of (solvent + salt) systems. The equilibrium data spanning vapor pressures of eighteen (solvent + salt) and three (solvent (1) + solvent (2) + salt) systems have been subjected to establish the basis for the model reliability analysis using a log-ratio objective function. The examined vapor pressure relations reproduce the observed performance relatively accurately, yielding the overall design factors of 1.084, 1.091 and 1.052 for the integrated property-basis solvation model (USMIP), reduced property-basis solvation model and concentration-dependent model, respectively. Both the integrated property-basis and reduced property-basis solvation models were able to simulate satisfactorily the vapor pressure data of a binary solvent mixture involving a salt, yielding an overall mean error of 5.2%

  1. Learning of spatio-temporal codes in a coupled oscillator system.

    Science.gov (United States)

    Orosz, Gábor; Ashwin, Peter; Townley, Stuart

    2009-07-01

    In this paper, we consider a learning strategy that allows one to transmit information between two coupled phase oscillator systems (called teaching and learning systems) via frequency adaptation. The dynamics of these systems can be modeled with reference to a number of partially synchronized cluster states and transitions between them. Forcing the teaching system by steady but spatially nonhomogeneous inputs produces cyclic sequences of transitions between the cluster states, that is, information about inputs is encoded via a "winnerless competition" process into spatio-temporal codes. The large variety of codes can be learned by the learning system that adapts its frequencies to those of the teaching system. We visualize the dynamics using "weighted order parameters (WOPs)" that are analogous to "local field potentials" in neural systems. Since spatio-temporal coding is a mechanism that appears in olfactory systems, the developed learning rules may help to extract information from these neural ensembles.

  2. Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

    Energy Technology Data Exchange (ETDEWEB)

    Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

    2017-04-01

    Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

  3. High throughput salt separation from uranium deposits

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, S.W.; Park, K.M.; Kim, J.G.; Kim, I.T.; Park, S.B., E-mail: swkwon@kaeri.re.kr [Korea Atomic Energy Research Inst. (Korea, Republic of)

    2014-07-01

    It is very important to increase the throughput of the salt separation system owing to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites in pyroprocessing. Multilayer porous crucible system was proposed to increase a throughput of the salt distiller in this study. An integrated sieve-crucible assembly was also investigated for the practical use of the porous crucible system. The salt evaporation behaviors were compared between the conventional nonporous crucible and the porous crucible. Two step weight reductions took place in the porous crucible, whereas the salt weight reduced only at high temperature by distillation in a nonporous crucible. The first weight reduction in the porous crucible was caused by the liquid salt penetrated out through the perforated crucible during the temperature elevation until the distillation temperature. Multilayer porous crucibles have a benefit to expand the evaporation surface area. (author)

  4. Phase Diagrams of the Aqueous Two-Phase Systems of Poly(ethylene glycol/Sodium Polyacrylate/Salts

    Directory of Open Access Journals (Sweden)

    Adalberto Pessoa Junior

    2011-03-01

    Full Text Available Aqueous two-phase systems consisting of polyethylene glycol (PEG, sodium polyacrylate (NaPAA, and a salt have been studied. The effects of the polymer size, salt type (NaCl, Na2SO4, sodium adipate and sodium azelate and salt concentrations on the position of the binodal curve were investigated. The investigated PEG molecules had a molar mass of 2,000 to 8,000 g/mol, while that of NaPAA was 8,000 g/mol. Experimental phase diagrams, and tie lines and calculated phase diagrams, based on Flory-Huggins theory of polymer solutions are presented. Due to strong enthalpic and entropic balancing forces, the hydrophobicity of the added salt has a strong influence on the position of the binodal, which could be reproduced by model calculations.

  5. Analysis of the KUCA MEU experiments using the ANL code system

    Energy Technology Data Exchange (ETDEWEB)

    Shiroya, S.; Hayashi, M.; Kanda, K.; Shibata, T.; Woodruff, W.L.; Matos, J.E.

    1982-01-01

    This paper provides some preliminary results on the analysis of the KUCA critical experiments using the ANL code system. Since this system was employed in the earlier neutronics calculations for the KUHFR, it is important to assess its capabilities for the KUHFR. The KUHFR has a unique core configuration which is difficult to model precisely with current diffusion theory codes. This paper also provides some results from a finite-element diffusion code (2D-FEM-KUR), which was developed in a cooperative research program between KURRI and JAERI. This code provides the capability for mockup of a complex core configuration as the KUHFR. Using the same group constants generated by the EPRI-CELL code, the results of the 2D-FEM-KUR code are compared with the finite difference diffusion code (DIF3D(2D) which is mainly employed in this analysis.

  6. Analysis of the KUCA MEU experiments using the ANL code system

    International Nuclear Information System (INIS)

    Shiroya, S.; Hayashi, M.; Kanda, K.; Shibata, T.; Woodruff, W.L.; Matos, J.E.

    1982-01-01

    This paper provides some preliminary results on the analysis of the KUCA critical experiments using the ANL code system. Since this system was employed in the earlier neutronics calculations for the KUHFR, it is important to assess its capabilities for the KUHFR. The KUHFR has a unique core configuration which is difficult to model precisely with current diffusion theory codes. This paper also provides some results from a finite-element diffusion code (2D-FEM-KUR), which was developed in a cooperative research program between KURRI and JAERI. This code provides the capability for mockup of a complex core configuration as the KUHFR. Using the same group constants generated by the EPRI-CELL code, the results of the 2D-FEM-KUR code are compared with the finite difference diffusion code (DIF3D(2D) which is mainly employed in this analysis

  7. Solubility diagrams in ternary aqueous-salt systems with yttrium, barium and copper nitrates at 25 deg C

    International Nuclear Information System (INIS)

    Petrenko, S.V.; Lyashchenko, A.K.; Karataeva, I.M.; Mozhaev, A.P.

    1990-01-01

    Solubility diagrams in Cu(NO 3 ) 2 -Y(NO 3 ) 3 -H 2 O, Y(NO 3 ) 3 -Ba(NO 3 ) 2 -H 2 O and Cu(NO 3 ) 2 -Ba(NO 3 ) 2 -H 2 O systems are studied at 25 deg C based on experimental determinations of salt solubility, isopiestic measurements and data of the calculation-experimental method in the framework of the Pitzer model. Binary and ternary parameters of interionic interactions are calculated; applicability of the method to the given systems is shown. Salts of complex composition and structure are absent in them; mutual salting out of components is also observed

  8. An experimental test plan for the characterization of molten salt thermochemical properties in heat transport systems

    International Nuclear Information System (INIS)

    Calderoni, Pattrick

    2010-01-01

    Molten salts are considered within the Very High Temperature Reactor program as heat transfer media because of their intrinsically favorable thermo-physical properties at temperatures starting from 300 C and extending up to 1200 C. In this context two main applications of molten salt are considered, both involving fluoride-based materials: as primary coolants for a heterogeneous fuel reactor core and as secondary heat transport medium to a helium power cycle for electricity generation or other processing plants, such as hydrogen production. The reference design concept here considered is the Advanced High Temperature Reactor (AHTR), which is a large passively safe reactor that uses solid graphite-matrix coated-particle fuel (similar to that used in gas-cooled reactors) and a molten salt primary and secondary coolant with peak temperatures between 700 and 1000 C, depending upon the application. However, the considerations included in this report apply to any high temperature system employing fluoride salts as heat transfer fluid, including intermediate heat exchangers for gas-cooled reactor concepts and homogeneous molten salt concepts, and extending also to fast reactors, accelerator-driven systems and fusion energy systems. The purpose of this report is to identify the technical issues related to the thermo-physical and thermo-chemical properties of the molten salts that would require experimental characterization in order to proceed with a credible design of heat transfer systems and their subsequent safety evaluation and licensing. In particular, the report outlines an experimental R and D test plan that would have to be incorporated as part of the design and operation of an engineering scaled facility aimed at validating molten salt heat transfer components, such as Intermediate Heat Exchangers. This report builds on a previous review of thermo-physical properties and thermo-chemical characteristics of candidate molten salt coolants that was generated as part

  9. An experimental test plan for the characterization of molten salt thermochemical properties in heat transport systems

    Energy Technology Data Exchange (ETDEWEB)

    Pattrick Calderoni

    2010-09-01

    Molten salts are considered within the Very High Temperature Reactor program as heat transfer media because of their intrinsically favorable thermo-physical properties at temperatures starting from 300 C and extending up to 1200 C. In this context two main applications of molten salt are considered, both involving fluoride-based materials: as primary coolants for a heterogeneous fuel reactor core and as secondary heat transport medium to a helium power cycle for electricity generation or other processing plants, such as hydrogen production. The reference design concept here considered is the Advanced High Temperature Reactor (AHTR), which is a large passively safe reactor that uses solid graphite-matrix coated-particle fuel (similar to that used in gas-cooled reactors) and a molten salt primary and secondary coolant with peak temperatures between 700 and 1000 C, depending upon the application. However, the considerations included in this report apply to any high temperature system employing fluoride salts as heat transfer fluid, including intermediate heat exchangers for gas-cooled reactor concepts and homogenous molten salt concepts, and extending also to fast reactors, accelerator-driven systems and fusion energy systems. The purpose of this report is to identify the technical issues related to the thermo-physical and thermo-chemical properties of the molten salts that would require experimental characterization in order to proceed with a credible design of heat transfer systems and their subsequent safety evaluation and licensing. In particular, the report outlines an experimental R&D test plan that would have to be incorporated as part of the design and operation of an engineering scaled facility aimed at validating molten salt heat transfer components, such as Intermediate Heat Exchangers. This report builds on a previous review of thermo-physical properties and thermo-chemical characteristics of candidate molten salt coolants that was generated as part of the

  10. Development of the versatile reactor analysis code system, MARBLE2

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Jin, Tomoyuki; Hazama, Taira; Hirai, Yasushi

    2015-07-01

    The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added in MARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, gamma-ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability. (author)

  11. Evaluation of the DRAGON code for VHTR design analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-12

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.

  12. Evaluation of the DRAGON code for VHTR design analysis

    International Nuclear Information System (INIS)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-01

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR

  13. High-Speed Turbo-TCM-Coded Orthogonal Frequency-Division Multiplexing Ultra-Wideband Systems

    Directory of Open Access Journals (Sweden)

    Wang Yanxia

    2006-01-01

    Full Text Available One of the UWB proposals in the IEEE P802.15 WPAN project is to use a multiband orthogonal frequency-division multiplexing (OFDM system and punctured convolutional codes for UWB channels supporting a data rate up to 480 Mbps. In this paper, we improve the proposed system using turbo TCM with QAM constellation for higher data rate transmission. We construct a punctured parity-concatenated trellis codes, in which a TCM code is used as the inner code and a simple parity-check code is employed as the outer code. The result shows that the system can offer a much higher spectral efficiency, for example, 1.2 Gbps, which is 2.5 times higher than the proposed system. We identify several essential requirements to achieve the high rate transmission, for example, frequency and time diversity and multilevel error protection. Results are confirmed by density evolution.

  14. The research of atmospheric 2D optical PPM CDMA system with turbo coding

    Science.gov (United States)

    Zhou, Xiuli; Li, Zaoxia

    2007-11-01

    The atmospheric two-dimensional optical code-division multiple-access (CDMA) systems using pulse-position modulation (PPM) and Turbo-coded were presented. We analyzed the bit-error rate (BER) of the proposed system using pulse-position modulation (PPM) with considering the effects of the scintillation, avalanche photodiode noise, thermal noise, and multi-user interference. We showed that the atmospheric two dimensional (2D) optical PPM CDMA systems can realize high-speed communications when the logarithm variance of the scintillation is less than 0.1, and the turbo-coded atmospheric optical CDMA system has better bit error rate(BER) performance than the atmospheric optical PPM CDMA systems without turbo-coded. We also showed that the turbo-coded system has better performance than the multi-user detection system.

  15. A new two dimensional spectral/spatial multi-diagonal code for noncoherent optical code division multiple access (OCDMA) systems

    Science.gov (United States)

    Kadhim, Rasim Azeez; Fadhil, Hilal Adnan; Aljunid, S. A.; Razalli, Mohamad Shahrazel

    2014-10-01

    A new two dimensional codes family, namely two dimensional multi-diagonal (2D-MD) codes, is proposed for spectral/spatial non-coherent OCDMA systems based on the one dimensional MD code. Since the MD code has the property of zero cross correlation, the proposed 2D-MD code also has this property. So that, the multi-access interference (MAI) is fully eliminated and the phase induced intensity noise (PIIN) is suppressed with the proposed code. Code performance is analyzed in terms of bit error rate (BER) while considering the effect of shot noise, PIIN, and thermal noise. The performance of the proposed code is compared with the related MD, modified quadratic congruence (MQC), two dimensional perfect difference (2D-PD) and two dimensional diluted perfect difference (2D-DPD) codes. The analytical and the simulation results reveal that the proposed 2D-MD code outperforms the other codes. Moreover, a large number of simultaneous users can be accommodated at low BER and high data rate.

  16. Innovative tank emptying system for the retrieval of salt, sludge and IX resins from storage tanks of NPPs

    International Nuclear Information System (INIS)

    Karl Froschauer; Holger Witing; Bernhard Christ

    2006-01-01

    RWE NUKEM recently developed a new Tank Emptying System (TESY) for the extraction of stored radioactive boric acid/borate salt blocks, sludge and IX resin from NPP stainless steel tanks of several hundred cubic meters content in Russia. RWE NUKEM has chosen the emptying concept consisting of a tracked submersible vehicle ('Crawler'), with jet nozzles for solution, agitation and fluidization, and a suction head to pick up the generated solution or suspension respectively. With the employment of RWE NUKEM's TESY system, spent radioactive salt deposits, ion-exchange resins and sludge, can be emptied and transferred out of the tank. The sediment, crystallized and settled during storage, will be agitated with increased temperature and suitable pH value and then picked up in form of a suspension or solution directly at the point of mobilization. This new Tank Emptying System concept enables efficiently to retrieve stored salt and other sediment waste, reduces operating time, safes cost for spare parts, increases the safety of operation and minimizes radiation exposure to personnel. All emptying tasks are performed remotely from a panel board and TV monitor located in a central control room. The TESY system consists of the following main components: glove box, crawler, submersible pump, heater, TV camera and spot light, control panel and monitor, water separation and feed unit, sodium hydroxide dosing unit. The system is specially requested for the removal of more than 2,500 cubic meter salt solution generated from the dissolution of some 300 cubic meter crystallized salt deposit per tank and per year. The TESY system is able to dissolve efficiently the salts and retrieve solutions and other liquefied suspensions. TESY is adaptable to all liquid waste storage facilities and especially deployable for tanks with limited access openings (<550 mm)

  17. Residual salts separation from metal reduced electrolytically in a LiCl-Li2O molten salt

    International Nuclear Information System (INIS)

    Hur, Jin Mok; Oh, Seung Chul; Hong, Sun Seok; Seo, Chung Seok; Park, Seong Won

    2005-01-01

    The PWR spent oxide fuel can be reduced electrolytically in a hot molten salt for the conditioning and the preparation of a metallic fuel. Then the metal product is smelted into an ingot to be treated in the post process. Incidentally, the residual salt which originated from the molten salt and spent fuel elements should be separated from the metal product during the smelting. In this work, we constructed a surrogate material system to simulate the salt separation from the reduced spent fuel and studied the vaporization behaviors of the salts

  18. Disposition of the fluoride fuel and flush salts from the Molten Salt Reactor experiment at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Peretz, F.J.

    1996-01-01

    The Molten Salt Reactor Experiment (MSRE) is an 8 MW reactor that was operated at Oak Ridge National Laboratory (ORNL) from 1965 through 1969. The reactor used a unique liquid salt fuel, composed of a mixture of LIF, BeF 2 , ZrF 4 , and UF 4 , and operated at temperatures above 600 degrees C. The primary fuel salt circulation system consisted of the reactor vessel, a single fuel salt pump, and a single primary heat exchanger. Heat was transferred from the fuel salt to a coolant salt circuit in the primary heat exchanger. The coolant salt was similar to the fuel salt, except that it contains only LiF (66%) and BeF, (34%). The coolant salt passed from the primary heat exchanger to an air-cooled radiator and a coolant salt pump, and then returned to the primary heat exchanger. Each of the salt loops was provided with drain tanks, located such that the salt could be drained out of either circuit by gravity. A single drain tank was provided for the non-radioactive coolant salt. Two drain tanks were provided for the fuel salt. Since the fuel salt contained radioactive fuel, fission products, and activation products, and since the reactor was designed such that the fuel salt could be drained immediately into the drain tanks in the event of a problem in the fuel salt loop, the fuel salt drain tanks were provided with a system to remove the heat generated by radioactive decay. A third drain tank connected to the fuel salt loop was provided for a batch of flush salt. This batch of salt, similar in composition to the coolant salt, was used to condition the fuel salt loop after it had been exposed to air and to flush the fuel salt loop of residual fuel salt prior to accessing the reactor circuit for maintenance or experimental activities. This report discusses the disposition of the fluoride fuel and flush salt

  19. Cardinality enhancement utilizing Sequential Algorithm (SeQ code in OCDMA system

    Directory of Open Access Journals (Sweden)

    Fazlina C. A. S.

    2017-01-01

    Full Text Available Optical Code Division Multiple Access (OCDMA has been important with increasing demand for high capacity and speed for communication in optical networks because of OCDMA technique high efficiency that can be achieved, hence fibre bandwidth is fully used. In this paper we will focus on Sequential Algorithm (SeQ code with AND detection technique using Optisystem design tool. The result revealed SeQ code capable to eliminate Multiple Access Interference (MAI and improve Bit Error Rate (BER, Phase Induced Intensity Noise (PIIN and orthogonally between users in the system. From the results, SeQ shows good performance of BER and capable to accommodate 190 numbers of simultaneous users contrast with existing code. Thus, SeQ code have enhanced the system about 36% and 111% of FCC and DCS code. In addition, SeQ have good BER performance 10-25 at 155 Mbps in comparison with 622 Mbps, 1 Gbps and 2 Gbps bit rate. From the plot graph, 155 Mbps bit rate is suitable enough speed for FTTH and LAN networks. Resolution can be made based on the superior performance of SeQ code. Thus, these codes will give an opportunity in OCDMA system for better quality of service in an optical access network for future generation's usage

  20. Cardinality enhancement utilizing Sequential Algorithm (SeQ) code in OCDMA system

    Science.gov (United States)

    Fazlina, C. A. S.; Rashidi, C. B. M.; Rahman, A. K.; Aljunid, S. A.

    2017-11-01

    Optical Code Division Multiple Access (OCDMA) has been important with increasing demand for high capacity and speed for communication in optical networks because of OCDMA technique high efficiency that can be achieved, hence fibre bandwidth is fully used. In this paper we will focus on Sequential Algorithm (SeQ) code with AND detection technique using Optisystem design tool. The result revealed SeQ code capable to eliminate Multiple Access Interference (MAI) and improve Bit Error Rate (BER), Phase Induced Intensity Noise (PIIN) and orthogonally between users in the system. From the results, SeQ shows good performance of BER and capable to accommodate 190 numbers of simultaneous users contrast with existing code. Thus, SeQ code have enhanced the system about 36% and 111% of FCC and DCS code. In addition, SeQ have good BER performance 10-25 at 155 Mbps in comparison with 622 Mbps, 1 Gbps and 2 Gbps bit rate. From the plot graph, 155 Mbps bit rate is suitable enough speed for FTTH and LAN networks. Resolution can be made based on the superior performance of SeQ code. Thus, these codes will give an opportunity in OCDMA system for better quality of service in an optical access network for future generation's usage

  1. Adaptive Wavelet Coding Applied in a Wireless Control System.

    Science.gov (United States)

    Gama, Felipe O S; Silveira, Luiz F Q; Salazar, Andrés O

    2017-12-13

    Wireless control systems can sense, control and act on the information exchanged between the wireless sensor nodes in a control loop. However, the exchanged information becomes susceptible to the degenerative effects produced by the multipath propagation. In order to minimize the destructive effects characteristic of wireless channels, several techniques have been investigated recently. Among them, wavelet coding is a good alternative for wireless communications for its robustness to the effects of multipath and its low computational complexity. This work proposes an adaptive wavelet coding whose parameters of code rate and signal constellation can vary according to the fading level and evaluates the use of this transmission system in a control loop implemented by wireless sensor nodes. The performance of the adaptive system was evaluated in terms of bit error rate (BER) versus E b / N 0 and spectral efficiency, considering a time-varying channel with flat Rayleigh fading, and in terms of processing overhead on a control system with wireless communication. The results obtained through computational simulations and experimental tests show performance gains obtained by insertion of the adaptive wavelet coding in a control loop with nodes interconnected by wireless link. These results enable the use of this technique in a wireless link control loop.

  2. Adaptive Wavelet Coding Applied in a Wireless Control System

    Directory of Open Access Journals (Sweden)

    Felipe O. S. Gama

    2017-12-01

    Full Text Available Wireless control systems can sense, control and act on the information exchanged between the wireless sensor nodes in a control loop. However, the exchanged information becomes susceptible to the degenerative effects produced by the multipath propagation. In order to minimize the destructive effects characteristic of wireless channels, several techniques have been investigated recently. Among them, wavelet coding is a good alternative for wireless communications for its robustness to the effects of multipath and its low computational complexity. This work proposes an adaptive wavelet coding whose parameters of code rate and signal constellation can vary according to the fading level and evaluates the use of this transmission system in a control loop implemented by wireless sensor nodes. The performance of the adaptive system was evaluated in terms of bit error rate (BER versus E b / N 0 and spectral efficiency, considering a time-varying channel with flat Rayleigh fading, and in terms of processing overhead on a control system with wireless communication. The results obtained through computational simulations and experimental tests show performance gains obtained by insertion of the adaptive wavelet coding in a control loop with nodes interconnected by wireless link. These results enable the use of this technique in a wireless link control loop.

  3. Code for calculation of spreading of radioactivity in reactor containment systems

    International Nuclear Information System (INIS)

    Vertes, P.

    1992-09-01

    A detailed description of the new version of TIBSO code is given, with applications for accident analysis in a reactor containment system. The TIBSO code can follow the nuclear transition and the spatial migration of radioactive materials. The modelling of such processes is established in a very flexible way enabling the user to investigate a wide range of problems. The TIBSO code system is described in detail, taking into account the new developments since 1983. Most changes improve the capabilities of the code. The new version of TIBSO system is written in FORTRAN-77 and can be operated both under VAX VMS and PC DOS. (author) 5 refs.; 3 figs.; 21 tabs

  4. The computer code system for reactor radiation shielding in design of nuclear power plant

    International Nuclear Information System (INIS)

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

    The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

  5. LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J M; Ahnert, C; Gomez Santamaria, J; Rodriguez Olabarria, I

    1985-07-01

    Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs.

  6. LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code

    International Nuclear Information System (INIS)

    Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.

    1985-01-01

    Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs

  7. Dynamic detection technology of malicious code for Android system

    Directory of Open Access Journals (Sweden)

    Li Boya

    2017-02-01

    Full Text Available With the increasing popularization of mobile phones,people's dependence on them is rising,the security problems become more and more prominent.According to the calling of the APK file permission and the API function in Android system,this paper proposes a dynamic detecting method based on API interception technology to detect the malicious code.The experimental results show that this method can effectively detect the malicious code in Android system.

  8. User effects on the transient system code calculations. Final report

    International Nuclear Information System (INIS)

    Aksan, S.N.; D'Auria, F.

    1995-01-01

    Large thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants to optimize operational procedures and the plant design itself. Evaluation of the capabilities of these codes are accomplished by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. In recent years, some attempts have been made to establish methodologies to evaluate the accuracy and the uncertainty of the code predictions and consequently judgement on the acceptability of the codes. In none of the methodologies has the influence of the code user on the calculated results been directly addressed. In this paper, the results of the investigations on the user effects for the thermal-hydraulic transient system codes is presented and discussed on the basis of some case studies. The general findings of the investigations show that in addition to user effects, there are other reasons that affect the results of the calculations and which are hidden under user effects. Both the hidden factors and the direct user effects are discussed in detail and general recommendations and conclusions are presented to control and limit them

  9. A comparative evaluation of NDR and PSAR using the CASMO-3/MASTER code system

    International Nuclear Information System (INIS)

    Sim, Jeoung Hun; Kim, Han Gon

    2009-01-01

    In order to validate nuclear design data such as the nuclear design report (NDR) and data in preliminary (or final) safety analysis report (PSAR/FSAR) and to use data for the conceptual design of new plants, the CASMO-3/MASTER code system is selected as utility code. The nuclear design of OPR1000 and APR1400 is performed with the DIT/ROCS code system. In contrast with this design code system, the accuracy of CASMO- 3/MASTER code system has not been verified. Relatively little design data has been calculated by the CASMO-3/MASTER code system for OPR1000 and APR1400 and a bias system has not been developed yet. As such, validation of the performance of the CASMO- 3/MASTER code system is necessary. In order to validate the performance of the CASMO- 3/MASTER code system and to develop a calculation methodology, a comparative evaluation with NDR of Ulchin unit 4, cycle 1(U4C1) and the PSAR of Shinkori units 3 and 4 is carried out. The results of this evaluation are presented in this paper

  10. Maximum Likelihood Blind Channel Estimation for Space-Time Coding Systems

    Directory of Open Access Journals (Sweden)

    Hakan A. Çırpan

    2002-05-01

    Full Text Available Sophisticated signal processing techniques have to be developed for capacity enhancement of future wireless communication systems. In recent years, space-time coding is proposed to provide significant capacity gains over the traditional communication systems in fading wireless channels. Space-time codes are obtained by combining channel coding, modulation, transmit diversity, and optional receive diversity in order to provide diversity at the receiver and coding gain without sacrificing the bandwidth. In this paper, we consider the problem of blind estimation of space-time coded signals along with the channel parameters. Both conditional and unconditional maximum likelihood approaches are developed and iterative solutions are proposed. The conditional maximum likelihood algorithm is based on iterative least squares with projection whereas the unconditional maximum likelihood approach is developed by means of finite state Markov process modelling. The performance analysis issues of the proposed methods are studied. Finally, some simulation results are presented.

  11. Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code

    Energy Technology Data Exchange (ETDEWEB)

    Guenay, Mehtap [Inoenue Univ., Malatya (Turkey). Physics Dept.

    2014-04-15

    In this study, the molten salt-heavy metal mixtures 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % NpO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion-fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library. (orig.)

  12. Design evaluation on sodium piping system and comparison of the design codes

    International Nuclear Information System (INIS)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon

    2015-01-01

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  13. Design evaluation on sodium piping system and comparison of the design codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon [KAERI, Daejeon (Korea, Republic of)

    2015-03-15

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  14. A systematic literature review of automated clinical coding and classification systems.

    Science.gov (United States)

    Stanfill, Mary H; Williams, Margaret; Fenton, Susan H; Jenders, Robert A; Hersh, William R

    2010-01-01

    Clinical coding and classification processes transform natural language descriptions in clinical text into data that can subsequently be used for clinical care, research, and other purposes. This systematic literature review examined studies that evaluated all types of automated coding and classification systems to determine the performance of such systems. Studies indexed in Medline or other relevant databases prior to March 2009 were considered. The 113 studies included in this review show that automated tools exist for a variety of coding and classification purposes, focus on various healthcare specialties, and handle a wide variety of clinical document types. Automated coding and classification systems themselves are not generalizable, nor are the results of the studies evaluating them. Published research shows these systems hold promise, but these data must be considered in context, with performance relative to the complexity of the task and the desired outcome.

  15. System Level Evaluation of Innovative Coded MIMO-OFDM Systems for Broadcasting Digital TV

    Directory of Open Access Journals (Sweden)

    Y. Nasser

    2008-01-01

    Full Text Available Single-frequency networks (SFNs for broadcasting digital TV is a topic of theoretical and practical interest for future broadcasting systems. Although progress has been made in the characterization of its description, there are still considerable gaps in its deployment with MIMO technique. The contribution of this paper is multifold. First, we investigate the possibility of applying a space-time (ST encoder between the antennas of two sites in SFN. Then, we introduce a 3D space-time-space block code for future terrestrial digital TV in SFN architecture. The proposed 3D code is based on a double-layer structure designed for intercell and intracell space time-coded transmissions. Eventually, we propose to adapt a technique called effective exponential signal-to-noise ratio (SNR mapping (EESM to predict the bit error rate (BER at the output of the channel decoder in the MIMO systems. The EESM technique as well as the simulations results will be used to doubly check the efficiency of our 3D code. This efficiency is obtained for equal and unequal received powers whatever is the location of the receiver by adequately combining ST codes. The 3D code is then a very promising candidate for SFN architecture with MIMO transmission.

  16. A bar-code reader for an alpha-beta automatic counting system - FAG

    Energy Technology Data Exchange (ETDEWEB)

    Levinson, S; Shemesh, Y; Ankry, N; Assido, H; German, U; Peled, O [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

    1996-12-01

    A bar-code laser system for sample number reading was integrated into the FAG Alpha-Beta automatic counting system. The sample identification by means of an attached bar-code label enables unmistakable and reliable attribution of results to the counted sample. Installation of the bar-code reader system required several modifications: Mechanical changes in the automatic sample changer, design and production of new sample holders, modification of the sample planchettes, changes in the electronic system, update of the operating software of the system (authors).

  17. A bar-code reader for an alpha-beta automatic counting system - FAG

    International Nuclear Information System (INIS)

    Levinson, S.; Shemesh, Y.; Ankry, N.; Assido, H.; German, U.; Peled, O.

    1996-01-01

    A bar-code laser system for sample number reading was integrated into the FAG Alpha-Beta automatic counting system. The sample identification by means of an attached bar-code label enables unmistakable and reliable attribution of results to the counted sample. Installation of the bar-code reader system required several modifications: Mechanical changes in the automatic sample changer, design and production of new sample holders, modification of the sample planchettes, changes in the electronic system, update of the operating software of the system (authors)

  18. Subchannel analysis of a boiloff experiment by a system thermalhydraulic code

    International Nuclear Information System (INIS)

    Bousbia-Salah, A.; D'Auria, F.

    2001-01-01

    This paper presents the results of system thermalhydraulic code using the sub-channel analysis approach in predicting the Neptun boil off experiments. This approach will be suitable for further works in view of coupling the system code with a 3D neutron kinetic one. The boil off tests were conducted in order to simulate the consequences of loss of coolant inventory leading to uncovery and heat up of fuel elements of a nuclear reactor core. In this framework, the Neptun low pressure test No5002, which is a good repeat experiment, is considered. The calculations were carried out using the system transient analysis code Relap5/Mod3.2. A detailed nodalization of the Neptun test section was developed. A reference case was run, and the overall data comparison shows good agreement between calculated and experimental thermalhydraulic parameters. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. The obtained results were almost satisfactory, this demonstrates, as well, the reasonable success of the subchannel analysis approach adopted in the present context for a system thermalhydraulic code.(author)

  19. Impact of Different Spreading Codes Using FEC on DWT Based MC-CDMA System

    OpenAIRE

    Masum, Saleh; Kabir, M. Hasnat; Islam, Md. Matiqul; Shams, Rifat Ara; Ullah, Shaikh Enayet

    2012-01-01

    The effect of different spreading codes in DWT based MC-CDMA wireless communication system is investigated. In this paper, we present the Bit Error Rate (BER) performance of different spreading codes (Walsh-Hadamard code, Orthogonal gold code and Golay complementary sequences) using Forward Error Correction (FEC) of the proposed system. The data is analyzed and is compared among different spreading codes in both coded and uncoded cases. It is found via computer simulation that the performance...

  20. Performance of asynchronous fiber-optic code division multiple access system based on three-dimensional wavelength/time/space codes and its link analysis.

    Science.gov (United States)

    Singh, Jaswinder

    2010-03-10

    A novel family of three-dimensional (3-D) wavelength/time/space codes for asynchronous optical code-division-multiple-access (CDMA) systems with "zero" off-peak autocorrelation and "unity" cross correlation is reported. Antipodal signaling and differential detection is employed in the system. A maximum of [(W x T+1) x W] codes are generated for unity cross correlation, where W and T are the number of wavelengths and time chips used in the code and are prime. The conditions for violation of the cross-correlation constraint are discussed. The expressions for number of generated codes are determined for various code dimensions. It is found that the maximum number of codes are generated for S systems. The codes have a code-set-size to code-size ratio greater than W/S. For instance, with a code size of 2065 (59 x 7 x 5), a total of 12,213 users can be supported, and 130 simultaneous users at a bit-error rate (BER) of 10(-9). An arrayed-waveguide-grating-based reconfigurable encoder/decoder design for 2-D implementation for the 3-D codes is presented so that the need for multiple star couplers and fiber ribbons is eliminated. The hardware requirements of the coders used for various modulation/detection schemes are given. The effect of insertion loss in the coders is shown to be significantly reduced with loss compensation by using an amplifier after encoding. An optical CDMA system for four users is simulated and the results presented show the improvement in performance with the use of loss compensation.

  1. Thorium cycle and molten salt reactors: field parameters and field constraints investigations toward 'thorium molten salt reactor' definition; Cycle thorium et reacteurs a sel fondu: exploration du champ des parametres et des contraintes definissant le 'Thorium Molten Salt Reactor'

    Energy Technology Data Exchange (ETDEWEB)

    Mathieu, L

    2005-09-15

    Producing nuclear energy in order to reduce the anthropic CO{sub 2} emission requires major technological advances. Nuclear plants of 4. generation have to respond to several constraints, as safety improvements, fuel breeding and radioactive waste minimization. For this purpose, it seems promising to use Thorium Cycle in Molten Salt Reactors. Studies on this domain have already been carried out. However, the final concept suffered from serious issues and was discontinued. A new reflection on this topic is being led in order to find acceptable solutions, and to design the Thorium Molten Salt Reactor concept. A nuclear reactor is simulated by the coupling of a neutron transport code with a materials evolution code. This allows us to reproduce the reactor behavior and its evolution all along its operation. Thanks to this method, we have studied a large number of reactor configurations. We have evaluated their efficiency through a group of constraints they have to satisfy. This work leads us to a better understanding of many physical phenomena controlling the reactor behavior. As a consequence, several efficient configurations have been discovered, allowing the emergence of new points of view in the research of Molten Salt Reactors. (author)

  2. OSCAR-4 Code System Application to the SAFARI-1 Reactor

    International Nuclear Information System (INIS)

    Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

    2008-01-01

    The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

  3. Java Source Code Analysis for API Migration to Embedded Systems

    Energy Technology Data Exchange (ETDEWEB)

    Winter, Victor [Univ. of Nebraska, Omaha, NE (United States); McCoy, James A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Guerrero, Jonathan [Univ. of Nebraska, Omaha, NE (United States); Reinke, Carl Werner [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Perry, James Thomas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-02-01

    Embedded systems form an integral part of our technological infrastructure and oftentimes play a complex and critical role within larger systems. From the perspective of reliability, security, and safety, strong arguments can be made favoring the use of Java over C in such systems. In part, this argument is based on the assumption that suitable subsets of Java’s APIs and extension libraries are available to embedded software developers. In practice, a number of Java-based embedded processors do not support the full features of the JVM. For such processors, source code migration is a mechanism by which key abstractions offered by APIs and extension libraries can made available to embedded software developers. The analysis required for Java source code-level library migration is based on the ability to correctly resolve element references to their corresponding element declarations. A key challenge in this setting is how to perform analysis for incomplete source-code bases (e.g., subsets of libraries) from which types and packages have been omitted. This article formalizes an approach that can be used to extend code bases targeted for migration in such a manner that the threats associated the analysis of incomplete code bases are eliminated.

  4. Nonterminals, homomorphisms and codings in different variations of OL-systems. II. Nondeterministic systems

    DEFF Research Database (Denmark)

    Nielsen, Mogens; Rozenberg, Grzegorz; Salomaa, Arto

    1974-01-01

    Continuing the work begun in Part I of this paper, we consider now variations of nondeterministic OL-systems. The present Part II of the paper contains a systematic classification of the effect of nonterminals, codings, weak codings, nonerasing homomorphisms and homomorphisms for all basic variat...

  5. Systemizers are better code-breakers:Self-reported systemizing predicts code-breaking performance in expert hackers and naïve participants

    Directory of Open Access Journals (Sweden)

    India eHarvey

    2016-05-01

    Full Text Available Studies on hacking have typically focused on motivational aspects and general personality traits of the individuals who engage in hacking; little systematic research has been conducted on predispositions that may be associated not only with the choice to pursue a hacking career but also with performance in either naïve or expert populations. Here we test the hypotheses that two traits that are typically enhanced in autism spectrum disorders - attention to detail and systemizing - may be positively related to both the choice of pursuing a career in information security and skilled performance in a prototypical hacking task (i.e. crypto-analysis or code-breaking. A group of naïve participants and of ethical hackers completed the Autism Spectrum Quotient, including an attention to detail scale, and the Systemizing Quotient (Baron-Cohen et al., 2001; Baron-Cohen et al., 2003. They were also tested with behavioural tasks involving code-breaking and a control task involving security x-ray image interpretation. Hackers reported significantly higher systemizing and attention to detail than non-hackers. We found a positive relation between self-reported systemizing (but not attention to detail and code-breaking skills in both hackers and non-hackers, whereas attention to detail (but not systemizing was related with performance in the x-ray screening task in both groups, as previously reported with naïve participants (Rusconi et al., 2015. We discuss the theoretical and translational implications of our findings.

  6. Systemizers Are Better Code-Breakers: Self-Reported Systemizing Predicts Code-Breaking Performance in Expert Hackers and Naïve Participants.

    Science.gov (United States)

    Harvey, India; Bolgan, Samuela; Mosca, Daniel; McLean, Colin; Rusconi, Elena

    2016-01-01

    Studies on hacking have typically focused on motivational aspects and general personality traits of the individuals who engage in hacking; little systematic research has been conducted on predispositions that may be associated not only with the choice to pursue a hacking career but also with performance in either naïve or expert populations. Here, we test the hypotheses that two traits that are typically enhanced in autism spectrum disorders-attention to detail and systemizing-may be positively related to both the choice of pursuing a career in information security and skilled performance in a prototypical hacking task (i.e., crypto-analysis or code-breaking). A group of naïve participants and of ethical hackers completed the Autism Spectrum Quotient, including an attention to detail scale, and the Systemizing Quotient (Baron-Cohen et al., 2001, 2003). They were also tested with behavioral tasks involving code-breaking and a control task involving security X-ray image interpretation. Hackers reported significantly higher systemizing and attention to detail than non-hackers. We found a positive relation between self-reported systemizing (but not attention to detail) and code-breaking skills in both hackers and non-hackers, whereas attention to detail (but not systemizing) was related with performance in the X-ray screening task in both groups, as previously reported with naïve participants (Rusconi et al., 2015). We discuss the theoretical and translational implications of our findings.

  7. Radioactivities evaluation code system for high temperature gas cooled reactors during normal operation

    International Nuclear Information System (INIS)

    Ogura, Kenji; Morimoto, Toshio; Suzuki, Katsuo.

    1979-01-01

    A radioactivity evaluation code system for high temperature gas-cooled reactors during normal operation was developed to study the behavior of fission products (FP) in the plants. The system consists of a code for the calculation of diffusion of FPs in fuel (FIPERX), a code for the deposition of FPs in primary cooling system (PLATO), a code for the transfer and emission of FPs in nuclear power plants (FIPPI-2), and a code for the exposure dose due to emitted FPs (FEDOSE). The FIPERX code can calculate the changes in the course of time FP of the distribution of FP concentration, the distribution of FP flow, the distribution of FP partial pressure, and the emission rate of FP into coolant. The amount of deposition of FPs and their distribution in primary cooling system can be evaluated by the PLATO code. The FIPPI-2 code can be used for the estimation of the amount of FPs in nuclear power plants and the amount of emitted FPs from the plants. The exposure dose of residents around nuclear power plants in case of the operation of the plants is calculated by the FEDOSE code. This code evaluates the dose due to the external exposure in the normal operation and in the accident, and the internal dose by the inhalation of radioactive plume and foods. Further studies of this code system by the comparison with the experimental data are considered. (Kato, T.)

  8. PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation

    International Nuclear Information System (INIS)

    Yaoqing, W.; Oppe, J.; Haas, J.B.M. de; Gruppelaar, H.; Slobben, J.

    1995-01-01

    1 - Description of program or function: The Petten AMPX-II/SCALE-3 Code System PASC-1 is a reactor neutronics calculation programme system consisting of well known IBM-oriented codes, that have been translated into FORTRAN-77, for calculations on a CDC-CYBER computer. Thus, the portability of these codes has been increased. In this system, some AMPX-II and SCALE-3 modules, the one-dimensional transport code ANISN and the 1 to 3-dimensional diffusion code CITATION are linked together on the CDC-CYBER/855 computer. The new cell code XSDRNPM-S and the old XSDRN code are included in the system. Starting from an AMPX fine group library up to CITATION, calculations can be performed for each individual module. Existing AMPX master interface format libraries, such as CSRL-IV, JEF-1, IRI and SCALE-45, and the old XSDRN-formatted libraries such as the COBB library can be used for the calculations. The code system contains the following modules and codes at present: AIM, AJAX, MALOCS, NITAWL-S, REVERT-I, ICE-2, CONVERT, JUAN, OCTAGN, XSDRNPM-S, XSDRN, ANISN and CITATION. The system will be extended with other SCALE modules and transport codes. 2 - Method of solution: The PASC-1 system is based on AMPX-II/SCALE-3 modules. Except for some SCALE-3 modules taken from the SCALIAS package, the original AMPX-II modules were IBM versions written in FORTRAN IV. These modules have been translated into CDC FORTRAN V. In order to test these modules and link them with some codes, some of the sample problem calculations have been performed for the whole PASC-1 system. During these calculations, some FORTRAN-77 errors were found in MALOCS, REVERT, CONVERT and some subroutines of SUBLIB (FORTRAN-77 subroutine library). These errors have been corrected. Because many corrections were made for the REVERT module, it is renamed as REVERT-I (improved version of REVERT). After these corrections, the whole system is running on a CDC-CYBER Computer (NOS-BE operating system). 3 - Restrictions on the

  9. Changes in the renin-angiotensin-aldosterone system in response to dietary salt intake in normal and hypertensive pregnancy. A randomized trial

    DEFF Research Database (Denmark)

    Nielsen, Lise Hald; Ovesen, Per; Hansen, Mie R

    2016-01-01

    It was hypothesized that primary renal sodium retention blunted the reactivity of the renin-angiotensin-aldosterone system to changes in salt intake in preeclampsia (PE). A randomized, cross-over, double-blinded, dietary intervention design was used to measure the effects of salt tablets or place...... of plasma renin and angiotensin II in response to changes in dietary salt intake compatible with a primary increase in renal sodium reabsorption in hypertensive pregnancies.......It was hypothesized that primary renal sodium retention blunted the reactivity of the renin-angiotensin-aldosterone system to changes in salt intake in preeclampsia (PE). A randomized, cross-over, double-blinded, dietary intervention design was used to measure the effects of salt tablets or placebo...

  10. Computational Analysis of Nanoparticles-Molten Salt Thermal Energy Storage for Concentrated Solar Power Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Vinod [Univ. of Texas, El Paso, TX (United States)

    2017-05-05

    High fidelity computational models of thermocline-based thermal energy storage (TES) were developed. The research goal was to advance the understanding of a single tank nanofludized molten salt based thermocline TES system under various concentration and sizes of the particles suspension. Our objectives were to utilize sensible-heat that operates with least irreversibility by using nanoscale physics. This was achieved by performing computational analysis of several storage designs, analyzing storage efficiency and estimating cost effectiveness for the TES systems under a concentrating solar power (CSP) scheme using molten salt as the storage medium. Since TES is one of the most costly but important components of a CSP plant, an efficient TES system has potential to make the electricity generated from solar technologies cost competitive with conventional sources of electricity.

  11. Diagonal Eigenvalue Unity (DEU) code for spectral amplitude coding-optical code division multiple access

    Science.gov (United States)

    Ahmed, Hassan Yousif; Nisar, K. S.

    2013-08-01

    Code with ideal in-phase cross correlation (CC) and practical code length to support high number of users are required in spectral amplitude coding-optical code division multiple access (SAC-OCDMA) systems. SAC systems are getting more attractive in the field of OCDMA because of its ability to eliminate the influence of multiple access interference (MAI) and also suppress the effect of phase induced intensity noise (PIIN). In this paper, we have proposed new Diagonal Eigenvalue Unity (DEU) code families with ideal in-phase CC based on Jordan block matrix with simple algebraic ways. Four sets of DEU code families based on the code weight W and number of users N for the combination (even, even), (even, odd), (odd, odd) and (odd, even) are constructed. This combination gives DEU code more flexibility in selection of code weight and number of users. These features made this code a compelling candidate for future optical communication systems. Numerical results show that the proposed DEU system outperforms reported codes. In addition, simulation results taken from a commercial optical systems simulator, Virtual Photonic Instrument (VPI™) shown that, using point to multipoint transmission in passive optical network (PON), DEU has better performance and could support long span with high data rate.

  12. Salt cookbook

    CERN Document Server

    Saha, Anirban

    2015-01-01

    If you are a professional associated with system and infrastructure management, looking at automated infrastructure and deployments, then this book is for you. No prior experience of Salt is required.

  13. Analog system for computing sparse codes

    Science.gov (United States)

    Rozell, Christopher John; Johnson, Don Herrick; Baraniuk, Richard Gordon; Olshausen, Bruno A.; Ortman, Robert Lowell

    2010-08-24

    A parallel dynamical system for computing sparse representations of data, i.e., where the data can be fully represented in terms of a small number of non-zero code elements, and for reconstructing compressively sensed images. The system is based on the principles of thresholding and local competition that solves a family of sparse approximation problems corresponding to various sparsity metrics. The system utilizes Locally Competitive Algorithms (LCAs), nodes in a population continually compete with neighboring units using (usually one-way) lateral inhibition to calculate coefficients representing an input in an over complete dictionary.

  14. APPC - A new standardised coding system for trans-organisational PACS retrieval

    International Nuclear Information System (INIS)

    Fruehwald, F.; Lindner, A.; Mostbeck, G.; Hruby, W.; Fruehwald-Pallamar, J.

    2010-01-01

    As part of a general strategy to integrate the health care enterprise, Austria plans to connect the Picture Archiving and Communication Systems (PACS) of all radiological institutions into a nationwide network. To facilitate the search for relevant correlative imaging data in the PACS of different organisations, a coding system was compiled for all radiological procedures and necessary anatomical details. This code, called the Austrian PACS Procedure Code (APPC), was granted the status of a standard under HL7. Examples are provided of effective coding and filtering when searching for relevant imaging material using the APPC, as well as the planned process for future adjustments of the APPC. The implementation and how the APPC will fit into the future electronic environment, which will include an electronic health act for all citizens in Austria, are discussed. A comparison to other nationwide electronic health record projects and coding systems is given. Limitations and possible use in physical storage media are contemplated. (orig.)

  15. Management-retrieval code system of fission barrier parameter sub-library

    International Nuclear Information System (INIS)

    Zhang Limin; Su Zongdi; Ge Zhigang

    1995-01-01

    The fission barrier parameter (FBP) library, which is a sub-library of Chinese Evaluated Nuclear Parameter library (CENPL), stores various popular used fission barrier parameters from different historical period, and could retrieve the required fission barrier parameters by using the management retrieval code system of the FBP sub-library. The function, feature and operation instruction of the code system are described briefly

  16. Fuel management and core design code systems for pressurized water reactor neutronic calculations

    International Nuclear Information System (INIS)

    Ahnert, C.; Arayones, J.M.

    1985-01-01

    A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

  17. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2004-02-01

    To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this

  18. Neutronic study of a nuclear reactor of fused salts

    International Nuclear Information System (INIS)

    Garcia B, F. B.; Francois L, J. L.

    2012-10-01

    The reactors of fused salts called Molten Salt Reactor have presented a resurgence of interest in the last decade, due to they have a versatility in particular to operate, either with a thermal or fast neutrons spectrum. The most active development was by the middle of 1950 and principles of 1970 in the Oak Ridge National Laboratory. In this work some developed models are presented particularly and studied with the help of the MCNPX code, for the development of the neutronic study of this reactor, starting of proposed models and from a simple and homogeneous geometry until other more complex models and approximate to more real cases. In particular the geometry conditions and criticality of each model were analyzed, the isotopic balance, as well as the concentrations of the salts and different assigned fuel types. (Author)

  19. Tritium module for ITER/Tiber system code

    International Nuclear Information System (INIS)

    Finn, P.A.; Willms, S.; Busigin, A.; Kalyanam, K.M.

    1988-01-01

    A tritium module was developed for the ITER/Tiber system code to provide information on capital costs, tritium inventory, power requirements and building volumes for these systems. In the tritium module, the main tritium subsystems/emdash/plasma processing, atmospheric cleanup, water cleanup, blanket processing/emdash/are each represented by simple scaleable algorithms. 6 refs., 2 tabs

  20. Analysis of PPM-CDMA and OPPM-CDMA communication systems with new optical code

    Science.gov (United States)

    Liu, F.; Ghafouri-Shiraz, H.

    2005-11-01

    A novel type of optical spreading sequences, named the 'new-Modified Prime Code (nMPC)', is proposed for use in synchronous direct-detection optical code-division multiple-access (CDMA) systems which employ both pulse position modulation (PPM) and overlapping pulse position modulation (OPPM) schemes. The upper bounds on the bit error rate (BER) for nMPC used in PPM-CDMA systems are derived and compared with the respective systems, using a modified prime code (MPC) and a padded modified prime code (PMPC). The nMPC is further applied to the OPPM-CDMA system and the system with a proposed interference cancellation scheme. Our results show that under the same conditions the PPM-CDMA system performances are more improved with the use of nMPC than with the two other traditional codes. Moreover, they show that the system performances are significantly enhanced by the proposed interference reduction methods, if the nMPC is used in the OPPM-CDMA systems.

  1. JAERI thermal reactor standard code system for reactor design and analysis SRAC

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro

    1985-01-01

    SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

  2. Coded aperture material motion detection system for the ACPR

    International Nuclear Information System (INIS)

    McArthur, D.A.; Kelly, J.G.

    1975-01-01

    Single LMFBR fuel pins are being irradiated in Sandia's Annular Core Pulsed Reactor (ACPR). In these experiments single fuel pins have been driven well into the melt and vaporization regions in transients with pulse widths of about 5 ms. The ACPR is being upgraded so that it can be used to irradiate bundles of seven LMFBR fuel pins. The coded aperture material motion detection system described is being developed for this upgraded ACPR, and has for its design goals 1 mm transverse resolution (i.e., in the axial and radial directions), depth resolution of a few cm, and time resolution of 0.1 ms. The target date for development of this system is fall 1977. The paper briefly reviews the properties of coded aperture imaging, describes one possible system for the ACPR upgrade, discusses experiments which have been performed to investigate the feasibility of such a system, and describes briefly the further work required to develop such a system. The type of coded aperture to be used has not yet been fixed, but a one-dimensional section of a Fresnel zone plate appears at this time to have significant advantages

  3. Coding Across Multicodes and Time in CDMA Systems Employing MMSE Multiuser Detector

    Directory of Open Access Journals (Sweden)

    Park Jeongsoon

    2004-01-01

    Full Text Available When combining a multicode CDMA system with convolutional coding, two methods have been considered in the literature. In one method, coding is across time in each multicode channel while in the other the coding is across both multicodes and time. In this paper, a performance/complexity analysis of decoding metrics and trellis structures for the two schemes is carried out. It is shown that the latter scheme can exploit the multicode diversity inherent in convolutionally coded direct sequence code division multiple access (DS-CDMA systems which employ minimum mean squared error (MMSE multiuser detectors. In particular, when the MMSE detector provides sufficiently different signal-to-interference ratios (SIRs for the multicode channels, coding across multicodes and time can obtain significant performance gain over coding across time, with nearly the same decoding complexity.

  4. Characterization of Salt Overly Sensitive 1 (SOS1) gene homoeologs in quinoa (Chenopodium quinoa Willd.).

    Science.gov (United States)

    Maughan, P J; Turner, T B; Coleman, C E; Elzinga, D B; Jellen, E N; Morales, J A; Udall, J A; Fairbanks, D J; Bonifacio, A

    2009-07-01

    Salt tolerance is an agronomically important trait that affects plant species around the globe. The Salt Overly Sensitive 1 (SOS1) gene encodes a plasma membrane Na+/H+ antiporter that plays an important role in germination and growth of plants in saline environments. Quinoa (Chenopodium quinoa Willd.) is a halophytic, allotetraploid grain crop of the family Amaranthaceae with impressive nutritional content and an increasing worldwide market. Many quinoa varieties have considerable salt tolerance, and research suggests quinoa may utilize novel mechanisms to confer salt tolerance. Here we report the cloning and characterization of two homoeologous SOS1 loci (cqSOS1A and cqSOS1B) from C. quinoa, including full-length cDNA sequences, genomic sequences, relative expression levels, fluorescent in situ hybridization (FISH) analysis, and a phylogenetic analysis of SOS1 genes from 13 plant taxa. The cqSOS1A and cqSOS1B genes each span 23 exons spread over 3477 bp and 3486 bp of coding sequence, respectively. These sequences share a high level of similarity with SOS1 homologs of other species and contain two conserved domains, a Nhap cation-antiporter domain and a cyclic-nucleotide binding domain. Genomic sequence analysis of two BAC clones (98 357 bp and 132 770 bp) containing the homoeologous SOS1 genes suggests possible conservation of synteny across the C. quinoa sub-genomes. This report represents the first molecular characterization of salt-tolerance genes in a halophytic species in the Amaranthaceae as well as the first comparative analysis of coding and non-coding DNA sequences of the two homoeologous genomes of C. quinoa.

  5. Improving system modeling accuracy with Monte Carlo codes

    International Nuclear Information System (INIS)

    Johnson, A.S.

    1996-01-01

    The use of computer codes based on Monte Carlo methods to perform criticality calculations has become common-place. Although results frequently published in the literature report calculated k eff values to four decimal places, people who use the codes in their everyday work say that they only believe the first two decimal places of any result. The lack of confidence in the computed k eff values may be due to the tendency of the reported standard deviation to underestimate errors associated with the Monte Carlo process. The standard deviation as reported by the codes is the standard deviation of the mean of the k eff values for individual generations in the computer simulation, not the standard deviation of the computed k eff value compared with the physical system. A more subtle problem with the standard deviation of the mean as reported by the codes is that all the k eff values from the separate generations are not statistically independent since the k eff of a given generation is a function of k eff of the previous generation, which is ultimately based on the starting source. To produce a standard deviation that is more representative of the physical system, statistically independent values of k eff are needed

  6. Stabilization through precipitation in a system of colloidal iron(III) pyrophosphate salts

    NARCIS (Netherlands)

    van Leeuwen, Y.M.; Velikov, K.P.; Kegel, W.K.

    2012-01-01

    The ionic strength of a solution decreases during the precipitation of an insoluble salt, which can cause an initially unstable colloidal system to stabilize during its formation. We show this effect in the precipitation and aging of colloidal iron(III) pyrophosphate, where we observe two distinct

  7. Thermodynamics investigation of a solar power system integrated oil and molten salt as heat transfer fluids

    International Nuclear Information System (INIS)

    Liu, Qibin; Bai, Zhang; Sun, Jie; Yan, Yuejun; Gao, Zhichao; Jin, Hongguang

    2016-01-01

    Highlights: • A new concentrating solar power system with a dual-solar field is proposed. • The superheated steam with more than 773 K is produced. • The performances of the proposed system are demonstrated. • The economic feasibility of the proposed system is validated. - Abstract: In this paper, a new parabolic trough solar power system that incorporates a dual-solar field with oil and molten salt as heat transfer fluids (HTFs) is proposed to effectively utilize the solar energy. The oil is chosen as a HTF in the low temperature solar field to heat the feeding water, and the high temperature solar field uses molten salt to superheat the steam that the temperature is higher than 773 K. The produced superheated steam enters a steam turbine to generate power. Energy analysis and exergy analysis of the system are implemented to evaluate the feasibility of the proposed system. Under considerations of variations of solar irradiation, the on-design and off-design thermodynamic performances of the system and the characteristics are investigated. The annual average solar-to-electric efficiency and the nominal efficiency under the given condition for the proposed solar thermal power generation system reach to 15.86% and 22.80%, which are higher than the reference system with a single HTF. The exergy losses within the solar heat transfer process of the proposed system are reduced by 7.8% and 45.23% compared with the solar power thermal systems using oil and molten salt as HTFs, respectively. The integrated approach with oil and molten salt as HTFs can make full use of the different physical properties of the HTFs, and optimize the heat transfer process between the HTFs and the water/steam. The exergy loss in the water evaporation and superheated process are reduced, the system efficiency and the economic performance are improved. The research findings provide a new approach for the improvement of the performances of solar thermal power plants.

  8. CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1989-02-01

    A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  9. Consumer acceptance of model soup system with varying levels of herbs and salt.

    Science.gov (United States)

    Wang, Chao; Lee, Youngsoo; Lee, Soo-Yeun

    2014-10-01

    Although herbs have been reported as one of the most common saltiness enhancers, few studies have focused on the effect of herbs on reducing added sodium as well as the impact of herbs on consumers' overall liking of foods. Therefore, the objectives of this study were to determine the effect of varying levels of herbs on reducing added sodium and consumers' overall liking of soups and identify the impact of salt levels on consumers' overall liking of soups. Overall liking of freshly prepared and retorted canned soups with varying levels of herbs was evaluated before and after adding salt by consumers ad libitum until the saltiness of the soup was just about right for them. The results of the study demonstrated that when the perceived herb flavor increased, the amount of salt consumers added to fresh soups decreased (P ≤ 0.006); however, consumers' overall liking decreased (P ≤ 0.013) as well for the highest level of herb tested in the study. Although overall liking of all canned soups was not significantly decreased by herbs, the amount of salt consumers added was also not significantly decreased when herbs were used. Overall liking of all soups significantly increased after more salt was added (P ≤ 0.001), which indicates that salt level was a dominant factor in affecting consumers' overall liking of soups with varying levels of herbs. These findings imply the role of herbs in decreasing salt intake, and the adequate amount of herbs to be added in soup systems. It is challenging for the food industry to reduce sodium in foods without fully understanding the impact of sodium reduction on sensory properties of foods. Herbs are recommended to use in reducing sodium; however, little has been reported regarding the effect of herbs on sodium reduction and how herbs influence consumers’ acceptance of foods. This study provides findings that herbs may aid in decreasing the amount of salt consumers need to add for freshly prepared soups. It was also found that high

  10. 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

  11. Thermal energy storage using chloride salts and their eutectics

    International Nuclear Information System (INIS)

    Myers, Philip D.; Goswami, D. Yogi

    2016-01-01

    Achieving the goals of the U.S. Department of Energy (DOE) Sunshot initiative requires (1) higher operating temperatures for concentrating solar power (CSP) plants to increase theoretical efficiency, and (2) effective thermal energy storage (TES) strategies to ensure dispatchability. Current inorganic salt-based TES systems in large-scale CSP plants generally employ molten nitrate salts for energy storage, but nitrate salts are limited in application to lower temperatures—generally, below 600 °C. These materials are sufficient for parabolic trough power plants, but they are inadequate for use at higher temperatures. At the higher operating temperatures achievable in solar power tower-type CSP plants, chloride salts are promising candidates for application as TES materials, owing to their thermal stability and generally lower cost compared to nitrate salts. In light of this, a recent study was conducted, which included a preliminary survey of chloride salts and binary eutectic systems that show promise as high temperature TES media. This study provided some basic information about the salts, including phase equilibria data and estimates of latent heat of fusion for some of the eutectics. Cost estimates were obtained through a review of bulk pricing for the pure salts among various vendors. This review paper updates that prior study, adding data for additional salt eutectic systems obtained from the literature. Where possible, data are obtained from the thermodynamic database software, FactSage. Radiative properties are presented, as well, since at higher temperatures, thermal radiation becomes a significant mode of heat transfer. Material compatibility for inorganic salts is another important consideration (e.g., with regard to piping and/or containment), so a summary of corrosion studies with various materials is also presented. Lastly, cost data for these systems are presented, allowing for meaningful comparison among these systems and other materials for TES

  12. A Nucleus-localized Long Non-Coding RNA Enhances Drought and Salt Stress Tolerance

    KAUST Repository

    Qin, Tao; Zhao, Huayan; Cui, Peng; Albesher, Nour H.; Xiong, Liming

    2017-01-01

    stress. DRIR was expressed at a low level under non-stress conditions but can be significantly activated by drought and salt stress as well as by abscisic acid (ABA) treatment. We identified a T-DNA insertion mutant, drirD, which had higher expression

  13. Symbol synchronization in convolutionally coded systems

    Science.gov (United States)

    Baumert, L. D.; Mceliece, R. J.; Van Tilborg, H. C. A.

    1979-01-01

    Alternate symbol inversion is sometimes applied to the output of convolutional encoders to guarantee sufficient richness of symbol transition for the receiver symbol synchronizer. A bound is given for the length of the transition-free symbol stream in such systems, and those convolutional codes are characterized in which arbitrarily long transition free runs occur.

  14. TOUGH+Hydrate v1.0 User's Manual: A Code for the Simulation of System Behavior in Hydrate-Bearing Geologic Media

    Energy Technology Data Exchange (ETDEWEB)

    Moridis, George; Moridis, George J.; Kowalsky, Michael B.; Pruess, Karsten

    2008-03-01

    TOUGH+HYDRATE v1.0 is a new code for the simulation of the behavior of hydrate-bearing geologic systems. By solving the coupled equations of mass and heat balance, TOUGH+HYDRATE can model the non-isothermal gas release, phase behavior and flow of fluids and heat under conditions typical of common natural CH{sub 4}-hydrate deposits (i.e., in the permafrost and in deep ocean sediments) in complex geological media at any scale (from laboratory to reservoir) at which Darcy's law is valid. TOUGH+HYDRATE v1.0 includes both an equilibrium and a kinetic model of hydrate formation and dissociation. The model accounts for heat and up to four mass components, i.e., water, CH{sub 4}, hydrate, and water-soluble inhibitors such as salts or alcohols. These are partitioned among four possible phases (gas phase, liquid phase, ice phase and hydrate phase). Hydrate dissociation or formation, phase changes and the corresponding thermal effects are fully described, as are the effects of inhibitors. The model can describe all possible hydrate dissociation mechanisms, i.e., depressurization, thermal stimulation, salting-out effects and inhibitor-induced effects. TOUGH+HYDRATE is the first member of TOUGH+, the successor to the TOUGH2 [Pruess et al., 1991] family of codes for multi-component, multiphase fluid and heat flow developed at the Lawrence Berkeley National Laboratory. It is written in standard FORTRAN 95, and can be run on any computational platform (workstation, PC, Macintosh) for which such compilers are available.

  15. High performance mixed optical CDMA system using ZCC code and multiband OFDM

    Directory of Open Access Journals (Sweden)

    Nawawi N. M.

    2017-01-01

    Full Text Available In this paper, we have proposed a high performance network design, which is based on mixed optical Code Division Multiple Access (CDMA system using Zero Cross Correlation (ZCC code and multiband Orthogonal Frequency Division Multiplexing (OFDM called catenated OFDM. In addition, we also investigate the related changing parameters such as; effective power, number of user, number of band, code length and code weight. Then we theoretically analyzed the system performance comprehensively while considering up to five OFDM bands. The feasibility of the proposed system architecture is verified via the numerical analysis. The research results demonstrated that our developed modulation solution can significantly enhanced the total number of user; improving up to 80% for five catenated bands compared to traditional optical CDMA system, with the code length equals to 80, transmitted at 622 Mbps. It is also demonstrated that the BER performance strongly depends on number of weight, especially with less number of users. As the number of weight increases, the BER performance is better.

  16. High performance mixed optical CDMA system using ZCC code and multiband OFDM

    Science.gov (United States)

    Nawawi, N. M.; Anuar, M. S.; Junita, M. N.; Rashidi, C. B. M.

    2017-11-01

    In this paper, we have proposed a high performance network design, which is based on mixed optical Code Division Multiple Access (CDMA) system using Zero Cross Correlation (ZCC) code and multiband Orthogonal Frequency Division Multiplexing (OFDM) called catenated OFDM. In addition, we also investigate the related changing parameters such as; effective power, number of user, number of band, code length and code weight. Then we theoretically analyzed the system performance comprehensively while considering up to five OFDM bands. The feasibility of the proposed system architecture is verified via the numerical analysis. The research results demonstrated that our developed modulation solution can significantly enhanced the total number of user; improving up to 80% for five catenated bands compared to traditional optical CDMA system, with the code length equals to 80, transmitted at 622 Mbps. It is also demonstrated that the BER performance strongly depends on number of weight, especially with less number of users. As the number of weight increases, the BER performance is better.

  17. The influence of bile salts on the distribution of simvastatin in the octanol/buffer system.

    Science.gov (United States)

    Đanić, Maja; Pavlović, Nebojša; Stanimirov, Bojan; Vukmirović, Saša; Nikolić, Katarina; Agbaba, Danica; Mikov, Momir

    2016-01-01

    Distribution coefficient (D) is useful parameter for evaluating drugs permeability properties across biological membranes, which are of importance for drugs bioavailability. Given that bile acids are intensively studied as drug permeation-modifying and -solubilizing agents, the aim of this study was to estimate the influence of sodium salts of cholic (CA), deoxycholic (DCA) and 12-monoketocholic acids (MKC) on distribution coefficient of simvastatin (SV) (lactone [SVL] and acid form [SVA]) which is a highly lipophilic compound with extremely low water solubility and bioavailability. LogD values of SVA and SVL with or without bile salts were measured by liquid-liquid extraction in n-octanol/buffer systems at pH 5 and 7.4. SV concentrations in aqueous phase were determined by HPLC-DAD. Chem3D Ultra program was applied for computation of physico-chemical properties of analyzed compounds and their complexes. Statistically significant decrease in both SVA and SVL logD was observed for all three studied bile salts at both selected pH. MKC exerted the most pronounced effect in the case of SVA while there were no statistically significant differences between observed bile salts for SVL. The calculated physico-chemical properties of analyzed compounds and their complexes supported experimental results. Our data indicate that the addition of bile salts into the n-octanol/buffer system decreases the values of SV distribution coefficient at both studied pH values. This may be the result of the formation of hydrophilic complexes increasing the solubility of SV that could consequently impact the pharmacokinetic parameters of SV and the final drug response in patients.

  18. On-line reprocessing of a molten salt reactor: a simulation tool

    International Nuclear Information System (INIS)

    Simon, Nicole; Gastaldi, Olivier; Penit, Thomas; Cohin, Olivier; Campion, Pierre-Yves

    2008-01-01

    The molten salt reactor (MSR) is one of the concepts studied in the frame of GEN IV road-map. Due to the specific features of its liquid fuel, the reprocessing unit may be directly connected to the reactor. A modelling of this unit is presented. The final objective is to create a flexible computer reprocessing code which can use data from neutron calculations and can be coupled to a neutron code. Such a code allows the description of the whole behaviour of MSR, including, in a coupled manner, both the design of the core and the optimised reprocessing scheme effects. (authors)

  19. Study of salinity in aqueous medium using X-Ray beam with MCNP-X code

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Caroline M.; Braz, Delson [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Salgado, César M., E-mail: cbarbosa@nuclear.ufrj.br, E-mail: delson@nuclear.ufrj.br, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    In offshore production, it is possible that the produced water presents geochemical characteristics that correspond to the mixture of formation water (connate water) and the sea water (injection water), and the physical-chemical behavior of the injected water allows a considerable variation in the index of salinity altering the water/oil ratio during transportation and/or extraction. Injection water is generally used to raise the reservoir pressure, increasing the percentage of extracted oil. This water has a significant amount of salts that generate some difficulties, such as measuring fractions of volume in multiphase systems. One way to check the effects of salinity would be to regularly measure the amount of salt present in the water. In this way, this work presents a methodology to measure the concentration and the types of salts using nuclear techniques through the MCNP-X computational code. The measurement geometry uses an X-ray beam (40-100 keV) and NaI(Tl) scintillation detector positioned diametrically opposed to the source. The studied samples were the NaCl, KCl and MgCl{sub 2} salts in aqueous solution. The results present the possibility of differentiating the formation and injection waters due to differences in the salt concentrations. (author)

  20. COUPLED SIMULATION OF GAS COOLED FAST REACTOR FUEL ASSEMBLY WITH NESTLE CODE SYSTEM

    Directory of Open Access Journals (Sweden)

    Filip Osusky

    2018-05-01

    Full Text Available The paper is focused on coupled calculation of the Gas Cooled Fast Reactor. The proper modelling of coupled neutronics and thermal-hydraulics is the corner stone for future safety assessment of the control and emergency systems. Nowadays, the system and channel thermal-hydraulic codes are accepted by the national regulatory authorities in European Union for license purposes, therefore the code NESTLE was used for the simulation. The NESTLE code is a coupled multigroup neutron diffusion code with thermal-hydraulic sub-channel code. In the paper, the validation of NESTLE code 5.2.1 installation is presented. The processing of fuel assembly homogeneous parametric cross-section library for NESTLE code simulation is made by the sequence TRITON of SCALE code package system. The simulated case in the NESTLE code is one fuel assembly of GFR2400 concept with reflective boundary condition in radial direction and zero flux boundary condition in axial direction. The results of coupled calculation are presented and are consistent with the GFR2400 study of the GoFastR project.

  1. Simulation of the containment spray system test PACOS PX2.2 with the integral code ASTEC and the containment code system COCOSYS

    International Nuclear Information System (INIS)

    Risken, Tobias; Koch, Marco K.

    2011-01-01

    The reactor safety research contains the analysis of postulated accidents in nuclear power plants (npp). These accidents may involve a loss of coolant from the nuclear plant's reactor coolant system, during which heat and pressure within the containment are increased. To handle these atmospheric conditions, containment spray systems are installed in various light water reactors (LWR) worldwide as a part of the accident management system. For the improvement and the safety ensurance in npp operation and accident management, numeric simulations of postulated accident scenarios are performed. The presented calculations regard the predictability of the containment spray system's effect with the integral code ASTEC and the containment code system COCOSYS, performed at Ruhr-Universitaet Bochum. Therefore the test PACOS Px2.2 is simulated, in which water is sprayed in the stratified containment atmosphere of the BMC (Battelle Modell-Containment). (orig.)

  2. Application bar-code system for solid radioactive waste management

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. H.; Kim, T. K.; Kang, I. S.; Cho, H. S.; Son, J. S. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    Solid radioactive wastes are generated from the post-irradiated fuel examination facility, the irradiated material examination facility, the research reactor, and the laboratories at KAERI. A bar-code system for a solid radioactive waste management of a research organization became necessary while developing the RAWMIS(Radioactive Waste Management Integration System) which it can generate personal history management for efficient management of a waste, documents, all kinds of statistics. This paper introduces an input and output application program design to do to database with data in the results and a stream process of a treatment that analyzed the waste occurrence present situation and data by bar-code system.

  3. A Novel Technique to Detect Code for SAC-OCDMA System

    Science.gov (United States)

    Bharti, Manisha; Kumar, Manoj; Sharma, Ajay K.

    2018-04-01

    The main task of optical code division multiple access (OCDMA) system is the detection of code used by a user in presence of multiple access interference (MAI). In this paper, new method of detection known as XOR subtraction detection for spectral amplitude coding OCDMA (SAC-OCDMA) based on double weight codes has been proposed and presented. As MAI is the main source of performance deterioration in OCDMA system, therefore, SAC technique is used in this paper to eliminate the effect of MAI up to a large extent. A comparative analysis is then made between the proposed scheme and other conventional detection schemes used like complimentary subtraction detection, AND subtraction detection and NAND subtraction detection. The system performance is characterized by Q-factor, BER and received optical power (ROP) with respect to input laser power and fiber length. The theoretical and simulation investigations reveal that the proposed detection technique provides better quality factor, security and received power in comparison to other conventional techniques. The wide opening of eye in case of proposed technique also proves its robustness.

  4. Systemizers Are Better Code-Breakers: Self-Reported Systemizing Predicts Code-Breaking Performance in Expert Hackers and Naïve Participants

    Science.gov (United States)

    Harvey, India; Bolgan, Samuela; Mosca, Daniel; McLean, Colin; Rusconi, Elena

    2016-01-01

    Studies on hacking have typically focused on motivational aspects and general personality traits of the individuals who engage in hacking; little systematic research has been conducted on predispositions that may be associated not only with the choice to pursue a hacking career but also with performance in either naïve or expert populations. Here, we test the hypotheses that two traits that are typically enhanced in autism spectrum disorders—attention to detail and systemizing—may be positively related to both the choice of pursuing a career in information security and skilled performance in a prototypical hacking task (i.e., crypto-analysis or code-breaking). A group of naïve participants and of ethical hackers completed the Autism Spectrum Quotient, including an attention to detail scale, and the Systemizing Quotient (Baron-Cohen et al., 2001, 2003). They were also tested with behavioral tasks involving code-breaking and a control task involving security X-ray image interpretation. Hackers reported significantly higher systemizing and attention to detail than non-hackers. We found a positive relation between self-reported systemizing (but not attention to detail) and code-breaking skills in both hackers and non-hackers, whereas attention to detail (but not systemizing) was related with performance in the X-ray screening task in both groups, as previously reported with naïve participants (Rusconi et al., 2015). We discuss the theoretical and translational implications of our findings. PMID:27242491

  5. Lithium isotope effect in the extraction systems of polyethers: effect of salt concentration

    International Nuclear Information System (INIS)

    Fang Shengqiang; Fu Lian

    1991-01-01

    Separation factors of lithium isotopes at 20 deg C were determined in the extraction systems of B15C5-CHCl 3 /LiBr-H 2 O. The initial concentration of LiBr was controlled in the extent of more than 2 mol/l. It may be established that the increase of LiBr concentration causes a remarkable increase of the separation factor. The essence of this effect due to the change in salt concentration was discussed in connection with examination of relevant phenomena in literature. It can be concluded that the relationship between α and Cm, the concentration of lithium salt, is dependent on K Q and K P express respectively, lithium isotope exchange equilibrium constants between Li-crownether complex and hydrated lithium ion for lithium concentration less than 1-2 mol/l, and between lithium salt ion pair and hydrated lithium ion for lithium concentration more than 2 mol/l in aqueous phase

  6. Progress on China nuclear data processing code system

    Science.gov (United States)

    Liu, Ping; Wu, Xiaofei; Ge, Zhigang; Li, Songyang; Wu, Haicheng; Wen, Lili; Wang, Wenming; Zhang, Huanyu

    2017-09-01

    China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.

  7. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    Energy Technology Data Exchange (ETDEWEB)

    Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)

    2015-09-30

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

  8. Coupled modelling of convergence, steel corrosion, gas production and brine flow in a rock salt repository

    International Nuclear Information System (INIS)

    Becker, D.A.; Hirsekorn, R.P.

    2013-01-01

    This poster presents the global simulation of the behaviour of thick-walled steel containers piled up in a borehole in a rock salt repository. The simulation takes into account: the convergence by the creeping of rock salt, the backfill and waste compaction, the porosity dependent flow resistance, the anaerobic corrosion (iron to magnetite transformation, gas production, brine consumption, water consumption and salt precipitation) and pressure development. Mechanical influence of corrosion has not yet been taken into account in the integrated code LOPOS

  9. Improving performance of DS-CDMA systems using chaotic complex Bernoulli spreading codes

    Science.gov (United States)

    Farzan Sabahi, Mohammad; Dehghanfard, Ali

    2014-12-01

    The most important goal of spreading spectrum communication system is to protect communication signals against interference and exploitation of information by unintended listeners. In fact, low probability of detection and low probability of intercept are two important parameters to increase the performance of the system. In Direct Sequence Code Division Multiple Access (DS-CDMA) systems, these properties are achieved by multiplying the data information in spreading sequences. Chaotic sequences, with their particular properties, have numerous applications in constructing spreading codes. Using one-dimensional Bernoulli chaotic sequence as spreading code is proposed in literature previously. The main feature of this sequence is its negative auto-correlation at lag of 1, which with proper design, leads to increase in efficiency of the communication system based on these codes. On the other hand, employing the complex chaotic sequences as spreading sequence also has been discussed in several papers. In this paper, use of two-dimensional Bernoulli chaotic sequences is proposed as spreading codes. The performance of a multi-user synchronous and asynchronous DS-CDMA system will be evaluated by applying these sequences under Additive White Gaussian Noise (AWGN) and fading channel. Simulation results indicate improvement of the performance in comparison with conventional spreading codes like Gold codes as well as similar complex chaotic spreading sequences. Similar to one-dimensional Bernoulli chaotic sequences, the proposed sequences also have negative auto-correlation. Besides, construction of complex sequences with lower average cross-correlation is possible with the proposed method.

  10. Variable weight Khazani-Syed code using hybrid fixed-dynamic technique for optical code division multiple access system

    Science.gov (United States)

    Anas, Siti Barirah Ahmad; Seyedzadeh, Saleh; Mokhtar, Makhfudzah; Sahbudin, Ratna Kalos Zakiah

    2016-10-01

    Future Internet consists of a wide spectrum of applications with different bit rates and quality of service (QoS) requirements. Prioritizing the services is essential to ensure that the delivery of information is at its best. Existing technologies have demonstrated how service differentiation techniques can be implemented in optical networks using data link and network layer operations. However, a physical layer approach can further improve system performance at a prescribed received signal quality by applying control at the bit level. This paper proposes a coding algorithm to support optical domain service differentiation using spectral amplitude coding techniques within an optical code division multiple access (OCDMA) scenario. A particular user or service has a varying weight applied to obtain the desired signal quality. The properties of the new code are compared with other OCDMA codes proposed for service differentiation. In addition, a mathematical model is developed for performance evaluation of the proposed code using two different detection techniques, namely direct decoding and complementary subtraction.

  11. Preliminary Study on the High Temperature Transport System for Molten Salt

    International Nuclear Information System (INIS)

    Lee, S. H.; Lee, H. S.; Kim, J. G.

    2012-01-01

    Pyroprocessing technology is one of the the most promising technologies for the advanced fuel cycle with favorable economic potential and intrinsic proliferation-resistance. The electrorefining process, one of main processes is compos- ed of pyroprocess to recover the useful elements from spent fuel, is under development at the Korea Atomic Energy Research Institute as a sub process of the pyrochemical treatment of spent PWR fuel. High-temperature molten salt transport technologies are required because a molten salt should be transported from the electrorefiner to electrowiner after the electrorefining process. Therefore, in pyroprocessing technology, the development of high-temperature transport technologies for molten salt is a crucial prerequisite. However, there have been a few transport studies on high-temperature molten salt. In this study, an apparatus for suction transport experiments was designed and constructed for the development of high temperature molten salt transport technology. Suction transport experiments were performed using LiC-KCl eutectic salt

  12. Restructuring of burnup sensitivity analysis code system by using an object-oriented design approach

    International Nuclear Information System (INIS)

    Kenji, Yokoyama; Makoto, Ishikawa; Masahiro, Tatsumi; Hideaki, Hyoudou

    2005-01-01

    A new burnup sensitivity analysis code system was developed with help from the object-oriented technique and written in Python language. It was confirmed that they are powerful to support complex numerical calculation procedure such as reactor burnup sensitivity analysis. The new burnup sensitivity analysis code system PSAGEP was restructured from a complicated old code system and reborn as a user-friendly code system which can calculate the sensitivity coefficients of the nuclear characteristics considering multicycle burnup effect based on the generalized perturbation theory (GPT). A new encapsulation framework for conventional codes written in Fortran was developed. This framework supported to restructure the software architecture of the old code system by hiding implementation details and allowed users of the new code system to easily calculate the burnup sensitivity coefficients. The framework can be applied to the other development projects since it is carefully designed to be independent from PSAGEP. Numerical results of the burnup sensitivity coefficient of a typical fast breeder reactor were given with components based on GPT and the multicycle burnup effects on the sensitivity coefficient were discussed. (authors)

  13. Feasibility analysis of the modified ATHLET code for supercritical water cooled systems

    Energy Technology Data Exchange (ETDEWEB)

    Zhou Chong, E-mail: ch.zhou@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany); Yang Yanhua [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Cheng Xu [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Modification of system code ATHLET for supercritical water application. Black-Right-Pointing-Pointer Development and assessment of a heat transfer package for supercritical water. Black-Right-Pointing-Pointer Validation of the modified code at supercritical pressures with the theoretical point-hydraulics model and the SASC code. Black-Right-Pointing-Pointer Application of the modified code to LOCA analysis of a supercritical water cooled in-pile fuel qualification test loop. - Abstract: Since the existing thermal-hydraulic computer codes for light water reactors are not applicable to supercritical water cooled reactors (SCWRs) owing to the limitation of physical models and numerical treatments, the development of a reliable thermal-hydraulic computer code is very important to design analysis and safety assessment of SCWRs. Based on earlier modification of ATHLET for SCWR, a general interface is implemented to the code, which serves as the platform for information exchange between ATHLET and the external independent physical modules. A heat transfer package containing five correlations for supercritical water is connected to the ATHLET code through the interface. The correlations are assessed with experimental data. To verify the modified ATHLET code, the Edwards-O'Brian blow-down test is simulated. As first validation at supercritical pressures, a simplified supercritical water cooled loop is modeled and its stability behavior is analyzed. Results are compared with that of the theoretical model and SASC code in the reference and show good agreement. To evaluate its feasibility, the modified ATHLET code is applied to a supercritical water cooled in-pile fuel qualification test loop. Loss of coolant accidents (LOCAs) due to break of coolant supply lines are calculated for the loop. Sensitivity analysis of some safety system parameters is performed to get further knowledge about their influence on the function of the

  14. Molten salts as possible fuel fluids for TRU fuelled systems: ISTC no. 1606 approach

    International Nuclear Information System (INIS)

    Ignatiev, V.; Zakirov, R.; Grebenkine, K.

    2001-01-01

    The principle attraction of the molten salt reactor (MSR) technology is the use of fuel/fertile material flexibility (easy of fuel preparation and processing) for gaining additional profits as compared with solid materials. This approach presents important departures from traditional philosophy, applied in current nuclear power plants, and to some extent contradicts the straightforward interpretation of the defence-in-depth principal. Nevertheless we understand there may be potential to use MSR technology to support back end fuel cycle technologies in future commercial environment. The paper aims at reviewing results of the work performed in Russia, relevant to the problems of MSR technology development. Also this contribution aims at evaluation of remaining uncertainties for molten salt burner concept implementation. Fuel properties and behaviour, container materials, and clean-up of fuels with emphasis on experiments will be of priority. Recommendations are made regarding the types of experimental studies needed on a way to implement molten salt technology to the back-end of the fuel cycle. To better understand the potential and limitations of the molten salts as a fuel for reactor of incinerator type, Russian Institutes have submitted to the ISTC the Task no. 1606 Experimental Study of Molten Salt Technology for Safe and Low Waste Treatment of Plutonium and Minor Actinides in Accelerator Driven and Critical Systems. The project goals, technical approach and expected specific results are discussed. (author)

  15. CENTAR code for extended nonlinear transient analysis of extraterrestrial reactor systems

    International Nuclear Information System (INIS)

    Nassersharif, B.; Peer, J.S.; DeHart, M.D.

    1987-01-01

    Current interest in the application of nuclear reactor-driven power systems to space missions has generated a need for a systems simulation code to model and analyze space reactor systems; such a code has been initiated at Texas A and M, and the first version is nearing completion; release was anticipated in the fall of 1987. This code, named CENTAR (Code for Extended Nonlinear Transient Analysis of Extraterrestrial Reactor Systems), is designed specifically for space systems and is highly vectorizable. CENTAR is composed of several specialized modules. A fluids module is used to model fluid behavior throughout the system. A wall heat transfer module models the heat transfer characteristics of all walls, insulation, and structure around the system. A fuel element thermal analysis module is used to predict the temperature behavior and heat transfer characteristics of the reactor fuel rods. A kinetics module uses a six-group point kinetics formulation to model reactivity feedback and control and the ANS 5.1 decay-heat curve to model shutdown decay-heat production. A pump module models the behavior of thermoelectric-electromagnetic pumps, and a heat exchanger module models not only thermal effects in thermoelectric heat exchangers, but also predicts electrical power production for a given configuration. Finally, an accumulator module models coolant expansion/contraction accumulators

  16. Analysis of an XADS Target with the System Code TRACE

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Sanchez Espinoza, Victor H.; Feng, Bo

    2008-01-01

    Accelerator-driven systems (ADS) present an option to reduce the radioactive waste of the nuclear industry. The experimental Accelerator-Driven System (XADS) has been designed to investigate the feasibility of using ADS on an industrial scale to burn minor actinides. The target section lies in the middle of the subcritical core and is bombarded by a proton beam to produce spallation neutrons. The thermal energy produced from this reaction requires a heat removal system for the target section. The target is cooled by liquid lead-bismuth-eutectics (LBE) in the primary system which in turn transfers the heat via a heat exchanger (HX) to the secondary coolant, Diphyl THT (DTHT), a synthetic diathermic fluid. Since this design is still in development, a detailed investigation of the system is necessary to evaluate the behavior during normal and transient operations. Due to the lack of experimental facilities and data for ADS, the analyses are mostly done using thermal hydraulic codes. In addition to evaluating the thermal hydraulics of the XADS, this paper also benchmarks a new code developed by the NRC, TRACE, against other established codes. The events used in this study are beam power switch-on/off transients and a loss of heat sink accident. The obtained results from TRACE were in good agreement with the results of various other codes. (authors)

  17. Development of the computer code system for the analyses of PWR core

    International Nuclear Information System (INIS)

    Tsujimoto, Iwao; Naito, Yoshitaka.

    1992-11-01

    This report is one of the materials for the work titled 'Development of the computer code system for the analyses of PWR core phenomena', which is performed under contracts between Shikoku Electric Power Company and JAERI. In this report, the numerical method adopted in our computer code system are described, that is, 'The basic course and the summary of the analysing method', 'Numerical method for solving the Boltzmann equation', 'Numerical method for solving the thermo-hydraulic equations' and 'Description on the computer code system'. (author)

  18. Home ranges of brown hares in a natural salt marsh: comparisons with agricultural systems

    NARCIS (Netherlands)

    Kunst, P.; Wal, van der R.; Wieren, van S.E.

    2001-01-01

    This is the first study on spatial behaviour of brown hares Lepus europaeus Pallas, 1778 based on radio-telemetry in a natural system, which we contrast with data from agricultural systems. Radio tracking took place in a Dutch salt marsh over a 10-month period, with intensive tracking sessions

  19. Home ranges of brown hares in a natural salt marsh : comparisons with agricultural systems

    NARCIS (Netherlands)

    Kunst, PJG; van der Wal, R; van Wieren, Sip

    This is the first study on spatial behaviour of brown hares Lepus europaeus Pallas, 1778 based on radio-telemetry in a natural system, which we contrast with data from agricultural systems. Radio tracking took place in a Dutch salt marsh over a 10-month period, with intensive tracking sessions

  20. On purpose simulation model for molten salt CSP parabolic trough

    Science.gov (United States)

    Caranese, Carlo; Matino, Francesca; Maccari, Augusto

    2017-06-01

    The utilization of computer codes and simulation software is one of the fundamental aspects for the development of any kind of technology and, in particular, in CSP sector for researchers, energy institutions, EPC and others stakeholders. In that extent, several models for the simulation of CSP plant have been developed with different main objectives (dynamic simulation, productivity analysis, techno economic optimization, etc.), each of which has shown its own validity and suitability. Some of those models have been designed to study several plant configurations taking into account different CSP plant technologies (Parabolic trough, Linear Fresnel, Solar Tower or Dish) and different settings for the heat transfer fluid, the thermal storage systems and for the overall plant operating logic. Due to a lack of direct experience of Molten Salt Parabolic Trough (MSPT) commercial plant operation, most of the simulation tools do not foresee a suitable management of the thermal energy storage logic and of the solar field freeze protection system, but follow standard schemes. ASSALT, Ase Software for SALT csp plants, has been developed to improve MSPT plant's simulations, by exploiting the most correct operational strategies in order to provide more accurate technical and economical results. In particular, ASSALT applies MSPT specific control logics for the electric energy production and delivery strategy as well as the operation modes of the Solar Field in off-normal sunshine condition. With this approach, the estimated plant efficiency is increased and the electricity consumptions required for the plant operation and management is drastically reduced. Here we present a first comparative study on a real case 55 MWe Molten Salt Parabolic Trough CSP plant placed in the Tibetan highlands, using ASSALT and SAM (System Advisor Model), which is a commercially available simulation tool.

  1. Library system for a one dimensional tokamak transport code: (LIBJT60), 1

    International Nuclear Information System (INIS)

    Hirayama, Toshio

    1982-12-01

    A library system is developed to control and manage huge programs in terms of FORTRAN source. It is applied to widely used one dimensional tokamak transport codes (LIBJT60), which have been developed in the Division of Large Tokamak Development. The structure of data and program in the transport code turn out to be flexible enough to respond to various demands and this gigantic code frame work can be decomposed into groups of a compact code with a specific function. Some editing support tools for programming and debugging are also developed to save programming work. By applying this library system, users can obtain a code whose functions can be efficiently developed. (author)

  2. Liquid salt environment stress-rupture testing

    Science.gov (United States)

    Ren, Weiju; Holcomb, David E.; Muralidharan, Govindarajan; Wilson, Dane F.

    2016-03-22

    Disclosed herein are systems, devices and methods for stress-rupture testing selected materials within a high-temperature liquid salt environment. Exemplary testing systems include a load train for holding a test specimen within a heated inert gas vessel. A thermal break included in the load train can thermally insulate a load cell positioned along the load train within the inert gas vessel. The test specimen can include a cylindrical gage portion having an internal void filled with a molten salt during stress-rupture testing. The gage portion can have an inner surface area to volume ratio of greater than 20 to maximize the corrosive effect of the molten salt on the specimen material during testing. Also disclosed are methods of making a salt ingot for placement within the test specimen.

  3. Transient and fuel performance analysis with VTT's coupled code system

    International Nuclear Information System (INIS)

    Daavittila, A.; Hamalainen, A.; Raty, H.

    2005-01-01

    VTT (technical research center of Finland) maintains and further develops a comprehensive safety analysis code system ranging from the basic neutronic libraries to 3-dimensional transient analysis and fuel behaviour analysis codes. The code system is based on various types of couplings between the relevant physical phenomena. The main tools for analyses of reactor transients are presently the 3-dimensional reactor dynamics code HEXTRAN for cores with a hexagonal fuel assembly geometry and TRAB-3D for cores with a quadratic fuel assembly geometry. HEXTRAN has been applied to safety analyses of VVER type reactors since early 1990's. TRAB-3D is the latest addition to the code system, and has been applied to BWR and PWR analyses in recent years. In this paper it is shown that TRAB-3D has calculated accurately the power distribution during the Olkiluoto-1 load rejection test. The results from the 3-dimensional analysis can be used as boundary conditions for more detailed fuel rod analysis. For this purpose a general flow model GENFLO, developed at VTT, has been coupled with USNRC's FRAPTRAN fuel accident behaviour model. The example case for FRAPTRAN-GENFLO is for an ATWS at a BWR plant. The basis for the analysis is an oscillation incident in the Olkiluoto-1 BWR during reactor startup on February 22, 1987. It is shown that the new coupled code FRAPTRAN/GENFLO is quite a promising tool that can handle flow situations and give a detailed analysis of reactor transients

  4. Scaling of Thermal-Hydraulic Phenomena and System Code Assessment

    International Nuclear Information System (INIS)

    Wolfert, K.

    2008-01-01

    In the last five decades large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Many separate effects tests and integral system tests were carried out to establish a data base for code development and code validation. In this context the question has to be answered, to what extent the results of down-scaled test facilities represent the thermal-hydraulic behaviour expected in a full-scale nuclear reactor under accidental conditions. Scaling principles, developed by many scientists and engineers, present a scientific technical basis and give a valuable orientation for the design of test facilities. However, it is impossible for a down-scaled facility to reproduce all physical phenomena in the correct temporal sequence and in the kind and strength of their occurrence. The designer needs to optimize a down-scaled facility for the processes of primary interest. This leads compulsorily to scaling distortions of other processes with less importance. Taking into account these weak points, a goal oriented code validation strategy is required, based on the analyses of separate effects tests and integral system tests as well as transients occurred in full-scale nuclear reactors. The CSNI validation matrices are an excellent basis for the fulfilling of this task. Separate effects tests in full scale play here an important role.

  5. Effect of salt stress on growth and physiology in amaranth and lettuce: Implications for bioregenerative life support system

    Science.gov (United States)

    Qin, Lifeng; Guo, Shuangsheng; Ai, Weidang; Tang, Yongkang; Cheng, Quanyong; Chen, Guang

    2013-02-01

    Growing plants can be used to clean waste water in bioregenerative life support system (BLSS). However, NaCl contained in the human urine always restricts plant growth and further reduces the degree of mass cycle closure of the system (i.e. salt stress). This work determined the effect of NaCl stress on physiological characteristics of plants for the life support system. Amaranth (Amaranthus tricolor L. var. Huahong) and leaf lettuce (Lactuca sativa L. var. Luoma) were cultivated at nutrient solutions with different NaCl contents (0, 1000, 5000 and 10,000 ppm, respectively) for 10 to 18 days after planted in the Controlled Ecological Life Support System Experimental Facility in China. Results showed that the two plants have different responses to the salt stress. The amaranth showed higher salt-tolerance with NaCl stress. If NaCl content in the solution is below 5000 ppm, the salt stress effect is insignificant on above-ground biomass output, leaf photosynthesis rate, Fv/Fm, photosynthesis pigment contents, activities of antioxidant enzymes, and inducing lipid peroxidation. On the other hand, the lettuce is sensitive to NaCl which significantly decreases those indices of growth and physiology. Notably, the lettuce remains high productivity of edible biomass in low NaCl stress, although its salt-tolerant limitation is lower than amaranth. Therefore, we recommended that amaranth could be cultivated under a higher NaCl stress condition (lettuce should be under a lower NaCl stress (<1000 ppm) for water cleaning in future BLSS.

  6. Effect of subsurface drainage on salt movement and distribution in salt-affected soils

    International Nuclear Information System (INIS)

    Moustafa, A.T.A.; Seliem, M.H.; Bakhati, H.K.

    1983-01-01

    This study was carried out to evaluate different subsurface drainage treatments (combinations of depth and spacing) on salt movement and distribution. The soil is clay and the drainage was designed according to the steady-state condition (Hooghoudt's equation). Three spacings and two depths resulted in six drainage treatments. Soil samples represented the initial state of every treatment and after 14 months they (cotton followed by wheat) were analysed. The data show that drain depth has its effective role in salt leaching, while drain spacing has its effect on salt distribution in the soil profile. The leaching rate of each specific ion is also affected by the different drainage treatments. In general, the salt movement and distribution should be taken into consideration when evaluating the design of drainage systems. (author)

  7. Secure Coding for Safety I and C Systems on Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Y. M.; Park, H. S.; Kim, T. H.

    2015-01-01

    This paper addresses secure coding technologies which can reduce the software vulnerabilities and provides secure coding application guidelines for nuclear safety I and C systems. The use of digital equipment may improve their reliability and reduce maintenance costs. But, the design characteristics of nuclear I and C systems are becoming more complex and the possibility of cyber-attacks using software vulnerabilities has been increased. Software defects, bugs and logic flaws have been consistently the primary causes of software vulnerabilities which can introduce security vulnerabilities. In this study, we described a applying methods for secure coding which can reduce the software vulnerabilities. Software defects lists, countermeasures for each defect and coding rules can be applied properly depending on target system's condition. We expect that the results of this study can help developing the secure coding guidelines and significantly reducing or eliminating vulnerabilities in nuclear safety I and C software

  8. Secure Coding for Safety I and C Systems on Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. M.; Park, H. S. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, T. H. [Formal Works Inc., Seoul (Korea, Republic of)

    2015-10-15

    This paper addresses secure coding technologies which can reduce the software vulnerabilities and provides secure coding application guidelines for nuclear safety I and C systems. The use of digital equipment may improve their reliability and reduce maintenance costs. But, the design characteristics of nuclear I and C systems are becoming more complex and the possibility of cyber-attacks using software vulnerabilities has been increased. Software defects, bugs and logic flaws have been consistently the primary causes of software vulnerabilities which can introduce security vulnerabilities. In this study, we described a applying methods for secure coding which can reduce the software vulnerabilities. Software defects lists, countermeasures for each defect and coding rules can be applied properly depending on target system's condition. We expect that the results of this study can help developing the secure coding guidelines and significantly reducing or eliminating vulnerabilities in nuclear safety I and C software.

  9. Regulatory monitoring systems of fortified salt and wheat flour in selected ASEAN countries.

    Science.gov (United States)

    van den Wijngaart, Annoek; Bégin, France; Codling, Karen; Randall, Philip; Johnson, Quentin W

    2013-06-01

    Considerable efforts have been made over the past decade to address vitamin and mineral deficiencies. An increasing number of countries in the Association of Southeast Asian Nations (ASEAN) are adopting mandatory food fortification as one of the primary strategies to overcome these deficiencies. Experience shows that fortified foods can reach large parts of the population, including the poor, if the fortification is done on a mandatory rather than a voluntary basis and if the food vehicle is widely consumed. To review the importance of regulatory monitoring as an essential component of food fortification efforts in selected ASEAN countries, with special focus on the available information on regulatory monitoring systems for iodized salt and fortified wheat flour. The role of regulatory monitoring in strengthening food fortification programs was discussed during a joint regional meeting of the World Health Organization, UNICEF, the Flour Fortification Initiative, the Global Alliance for Improved Nutrition, the Micronutrient Initiative, and the World Bank on regulatory monitoring of salt and wheat flour fortification programs in Asia, which took place in Manila, Philippines, on 27-29 September 2011. This paper reviews the regulatory monitoring systems of selected ASEAN countries that participated in this meeting. Problems and challenges in regulatory monitoring systems for iodized salt and fortified wheat flour in selected ASEAN countries are identified, and a description of the role of regulatory monitoring in strengthening food fortification initiatives, particularly of salt and flour, and highlights of areas for improvement are presented. Regulatory monitoring consists of monitoring activities conducted at the production level, at customs warehouses, and at retail stores by concerned regulatory authorities, and at the production level by producers themselves, as part of quality control and assurance efforts. Unless there are appropriate enforcement and quality

  10. Experimental studies on natural circulation in molten salt loops

    International Nuclear Information System (INIS)

    Srivastava, A.K.; Borgohain, A.; Maheshwari, N.K.; Vijayan, P.K.

    2015-01-01

    Molten salts are increasingly getting attention as a coolant and storage medium in solar thermal power plants and as a liquid fuel, blanket and coolant in Molten Salt Reactors (MSR’s). Two different test facilities named Molten Salt Natural Circulation Loop (MSNCL) and Molten Active Fluoride salt Loop (MAFL) have been setup for thermal hydraulics, instrument development and material related studies relevant to MSR and solar power plants. The working medium for MSNCL is a molten nitrate salt which is a mixture of NaNO 3 and KNO 3 in 60:40 ratio and proposed as one of the coolant option for molten salt based reactor and coolant as well as storage medium for solar thermal power application. On the other hand, the working medium for MAFL is a eutectic mixture of LiF and ThF 4 and proposed as a blanket salt for Indian Molten Salt Breeder Reactor (MSBR). Steady state natural circulation experiments at different power level have been performed in the MSNCL. Transient studies for startup of natural circulation, loss of heat sink, heater trip and step change in heater power have also been carried out in the same. A 1D code LeBENC, developed in-house to simulate the natural circulation characteristics in closed loops, has been validated with the experimental data obtained from MSNCL. Further, LeBENC has been used for Pretest analysis of MAFL. This paper deals with the description of both the loops and experimental studies carried out in MSNCL. Validation of LeBENC along with the pretest analysis of MAFL using the same are also reported in this paper. (author)

  11. Results from the coded aperture neutron imaging system

    International Nuclear Information System (INIS)

    Brubaker, Erik; Steele, John T.; Brennan, James S.; Marleau, Peter

    2010-01-01

    Because of their penetrating power, energetic neutrons and gamma rays (∼1 MeV) offer the best possibility of detecting highly shielded or distant special nuclear material (SNM). Of these, fast neutrons offer the greatest advantage due to their very low and well understood natural background. We are investigating a new approach to fast-neutron imaging - a coded aperture neutron imaging system (CANIS). Coded aperture neutron imaging should offer a highly efficient solution for improved detection speed, range, and sensitivity. We have demonstrated fast neutron and gamma ray imaging with several different configurations of coded masks patterns and detectors including an 'active' mask that is composed of neutron detectors. Here we describe our prototype detector and present some initial results from laboratory tests and demonstrations.

  12. Code-division multiple-access protocol for active RFID systems

    Science.gov (United States)

    Mazurek, Gustaw; Szabatin, Jerzy

    2008-01-01

    Most of the Radio Frequency Identification (RFID) systems operating in HF and UHF bands employ narrowband modulations (FSK or ASK) with Manchester coding. However, these simple transmission schemes are vulnerable to narrowband interference (NBI) generated by other radio systems working in the same frequency band, and also suffer from collision problem and need special anti-collision procedures. This becomes especially important when operating in a noisy, crowded industrial environment. In this paper we show the performance of RFID system with DS-CDMA transmission in comparison to a standard system with FSK modulation defined in ISO 18000-7. Our simulation results show that without any bandwidth expansion the immunity against NBI can be improved by 8 dB and the system capacity can be 7 times higher when using DS-CDMA transmission instead of FSK modulation with Manchester coding.

  13. Implantation of Magint-Maggraf code in the Microvax-3600 system

    International Nuclear Information System (INIS)

    Leite Neto, J.P.; Shiabata, C.S.; Montes, A.

    1990-01-01

    An auxiliary code, named MAGGRAF, was developed and implemented on the MicroVAX-3600 system of the Plasma Laboratory at the Institute for Space Research, in order to perform the graphical output of the numerical results provided by the magnetic field calculation code MAGINT. In this report we present a brief description of the graphical code, of the parameters which specify the different output options, and of the structure of the data file containing these parameters. Some examples are shown to illustrate the versatility of the code, as well as the quality of the graphs. (author)

  14. Reconciliation of international administrative coding systems for comparison of colorectal surgery outcome.

    Science.gov (United States)

    Munasinghe, A; Chang, D; Mamidanna, R; Middleton, S; Joy, M; Penninckx, F; Darzi, A; Livingston, E; Faiz, O

    2014-07-01

    Significant variation in colorectal surgery outcomes exists between different countries. Better understanding of the sources of variable outcomes using administrative data requires alignment of differing clinical coding systems. We aimed to map similar diagnoses and procedures across administrative coding systems used in different countries. Administrative data were collected in a central database as part of the Global Comparators (GC) Project. In order to unify these data, a systematic translation of diagnostic and procedural codes was undertaken. Codes for colorectal diagnoses, resections, operative complications and reoperative interventions were mapped across the respective national healthcare administrative coding systems. Discharge data from January 2006 to June 2011 for patients who had undergone colorectal surgical resections were analysed to generate risk-adjusted models for mortality, length of stay, readmissions and reoperations. In all, 52 544 case records were collated from 31 institutions in five countries. Mapping of all the coding systems was achieved so that diagnosis and procedures from the participant countries could be compared. Using the aligned coding systems to develop risk-adjusted models, the 30-day mortality rate for colorectal surgery was 3.95% (95% CI 0.86-7.54), the 30-day readmission rate was 11.05% (5.67-17.61), the 28-day reoperation rate was 6.13% (3.68-9.66) and the mean length of stay was 14 (7.65-46.76) days. The linkage of international hospital administrative data that we developed enabled comparison of documented surgical outcomes between countries. This methodology may facilitate international benchmarking. Colorectal Disease © 2014 The Association of Coloproctology of Great Britain and Ireland.

  15. Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

    International Nuclear Information System (INIS)

    Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang

    2005-01-01

    Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

  16. Investigating the sealing capacity of a seal system in rock salt (DOPAS project)

    Energy Technology Data Exchange (ETDEWEB)

    Jantschik, Kyra; Moog, Helge C.; Czaikowski, Oliver; Wieczorek, Klaus [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Braunschweig (Germany)

    2016-11-15

    This paper describes research and development work on plugging and sealing repositories, an issue of fundamental importance for the rock salt option which represents one of the three European repository options, besides the clay rock and the crystalline rock options. The programme aims at providing experimental data needed for the theoretical analysis of the long-term sealing capacity of concrete- based sealing materials. In order to demonstrate hydro-mechanical material stability under representative load scenarios, a comprehensive laboratory testing programme is carried out. This comprises investigation of the sealing capacity of the combined seal system and impact of the so-called excavation-damaged zones (EDZ) as well as investigation of the hydro-chemical long-term stability of the seal in contact with different brines under diffusive and advective conditions. This paper presents experimental approaches and preliminary results from laboratory investigations on salt concrete and combined systems as obtained to date.

  17. Liking, salt taste perception and use of table salt when consuming reduced-salt chicken stews in light of South Africa's new salt regulations.

    Science.gov (United States)

    De Kock, H L; Zandstra, E H; Sayed, N; Wentzel-Viljoen, E

    2016-01-01

    This study investigated the impact of salt reduction on liking, salt taste perception, and use of table salt when consuming chicken stew in light of South Africa's new salt recommendations. In total, 432 South-African consumers (aged 35.2 ± 12.3 years) consumed a full portion of a chicken stew meal once at a central location. Four stock cube powders varying in salt content were used to prepare chicken stews: 1) no reduction - 2013 Na level; regular salt level as currently available on the South African market (24473 mg Na/100 g), 2) salt reduction smaller than 2016 level, i.e. 10%-reduced (22025 mg Na/100 g), 3) 2016 salt level, as per regulatory prescriptions (18000 mg Na/100 g), 4) 2019 salt level, as per regulatory prescriptions (13000 mg Na/100 g). Consumers were randomly allocated to consume one of the four meals. Liking, salt taste perception, and use of table salt and pepper were measured. Chicken stews prepared with reduced-salt stock powders were equally well-liked as chicken stews with the current salt level. Moreover, a gradual reduction of the salt in the chicken stews resulted in a reduced salt intake, up to an average of 19% for the total group compared to the benchmark 2013 Na level stew. However, 19% of consumers compensated by adding salt back to full compensation in some cases. More salt was added with increased reductions of salt in the meals, even to the point of full compensation. Further investigation into the impacts of nutrition communication and education about salt reduction on salt taste perception and use is needed. This research provides new consumer insights on salt use and emphasises the need for consumer-focused behaviour change approaches, in addition to reformulation of products. Copyright © 2015 Elsevier Ltd. All rights reserved.

  18. Simulation of salt production process

    Science.gov (United States)

    Muraveva, E. A.

    2017-10-01

    In this paper an approach to the use of simulation software iThink to simulate the salt production system has been proposed. The dynamic processes of the original system are substituted by processes simulated in the abstract model, but in compliance with the basic rules of the original system, which allows one to accelerate and reduce the cost of the research. As a result, a stable workable simulation model was obtained that can display the rate of the salt exhaustion and many other parameters which are important for business planning.

  19. Sea Salt vs. Table Salt: What's the Difference?

    Science.gov (United States)

    ... and healthy eating What's the difference between sea salt and table salt? Answers from Katherine Zeratsky, R.D., L.D. The main differences between sea salt and table salt are in their taste, texture ...

  20. A ''SuperCode'' for performing systems analysis of tokamak experiments and reactors

    International Nuclear Information System (INIS)

    Haney, S.W.; Barr, W.L.; Crotinger, J.A.; Perkins, L.J.; Solomon, C.J.; Chaniotakis, E.A.; Freidberg, J.P.; Wei, J.; Galambos, J.D.; Mandrekas, J.

    1992-01-01

    A new code, named the ''SUPERCODE,'' has been developed to fill the gap between currently available zero dimensional systems codes and highly sophisticated, multidimensional plasma performance codes. The former are comprehensive in content, fast to execute, but rather simple in terms of the accuracy of the physics and engineering models. The latter contain state-of-the-art plasma physics modelling but are limited in engineering content and time consuming to run. The SUPERCODE upgrades the reliability and accuracy of systems codes by calculating the self consistent 1 1/2 dimensional MHD-transport plasma evolution in a realistic engineering environment. By a combination of variational techniques and careful formation, there is only a modest increase in CPU time over O-D runs, thereby making the SUPERCODE suitable for use as a systems studies tool. In addition, considerable effort has been expended to make the code user- and programming-friendly, as well as operationally flexible, with the hope of encouraging wide usage throughout the fusion community

  1. An Investigation on the Thermophysical Properties of a Binary Molten Salt System Containing Both Aluminum Oxide and Titanium Oxide Nanoparticle Suspensions

    Science.gov (United States)

    Giridhar, Kunal

    Molten salts are showing great potential to replace current heat transfer and thermal energy storage fluids in concentrated solar plants because of their capability to maximize thermal energy storage, greater stability, cost effectiveness and significant thermal properties. However one of the major drawbacks of using molten salt as heat transfer fluid is that they are in solid state at room temperature and they have a high freezing point. Hence, significant resources would be required to maintain it in liquid form. If molten salt freezes while in operation, it would eventually damage piping network due to its volume shrinkage along with rendering the entire plant inoperable. It is long known that addition of nanoparticle suspensions has led to significant changes in thermal properties of fluids. In this investigation, aluminum oxide and titanium oxide nanoparticles of varying concentrations are added to molten salt/solar salt system consisting of 60% sodium nitrate and 40% potassium nitrate. Using differential scanning calorimeter, an attempt will be made to investigate changes in heat capacity of system, depression in freezing point and changes in latent heat of fusion. Scanning electron microscope will be used to take images of samples to study changes in micro-structure of mixture, ensure uniform distribution of nanoparticle in system and verify authenticity of materials used for experimentation. Due to enormous magnitude of CSP plant, actual implementation of molten salt system is on a large scale. With this investigation, even microscopic enhancement in heat capacity and slight lowering of freezing point will lead to greater benefits in terms of efficiency and cost of operation of plant. These results will further the argument for viability of molten salt as a heat transfer fluid and thermal storage system in CSP. One of the objective of this experimentation is to also collect experimental data which can be used for establishing relation between concentration

  2. Front-end and back-end electrochemistry of molten salt in accelerator-driven transmutation systems

    International Nuclear Information System (INIS)

    Williamson, M.A.; Venneri, F.

    1995-01-01

    The objective of this work is to develop preparation and clean-up processes for the fuel and carrier salt in the Los Alamos Accelerator-Driven Transmutation Technology molten salt nuclear system. The front-end or fuel preparation process focuses on the removal of fission products, uranium, and zirconium from spent nuclear fuel by utilizing electrochemical methods (i.e., electrowinning). The same method provides the separation of the so-called noble metal fission products at the back-end of the fuel cycle. Both implementations would have important diversion safeguards. The proposed separation processes and a thermodynamic analysis of the electrochemical separation method are presented

  3. Performance analysis of conceptual waste package designs in salt repositories

    International Nuclear Information System (INIS)

    Jansen, G. Jr.; Raines, G.E.; Kircher, J.F.

    1984-01-01

    A performance analysis of commercial high-level waste and spent fuel conceptual package designs in reference repositories in three salt formations was conducted with the WAPPA waste package code. Expected conditions for temperature, stress, brine composition, radiation level, and brine flow rate were used as boundary conditions to compute expected corrosion of a thick-walled overpack of 1025 wrought steel. In all salt formations corrosion by low Mg salt-dissolution brines typical of intrusion scenarios was too slow to cause the package to fail for thousands of years after burial. In high Mg brines judged typical of thermally migrating brines in bedded salt formations, corrosion rates which would otherwise have caused the packages to fail within a few hundred years were limited by brine availability. All of the brine reaching the package was consumed by reaction with the iron in the overpack, thus preventing further corrosion. Uniform brine distribution over the package surface was an important factor in predicting long package lifetimes for the high Mg brines. 14 references, 15 figures

  4. High-salt diets during pregnancy affected fetal and offspring renal renin-angiotensin system.

    Science.gov (United States)

    Mao, Caiping; Liu, Rong; Bo, Le; Chen, Ningjing; Li, Shigang; Xia, Shuixiu; Chen, Jie; Li, Dawei; Zhang, Lubo; Xu, Zhice

    2013-07-01

    Intrauterine environments are related to fetal renal development and postnatal health. Influence of salty diets during pregnancy on renal functions and renin-angiotensin system (RAS) was determined in the ovine fetuses and offspring. Pregnant ewes were fed high-salt diet (HSD) or normal-salt diet (NSD) for 2 months during middle-to-late gestation. Fetal renal functions, plasma hormones, and mRNA and protein expressions of the key elements of renal RAS were measured in the fetuses and offspring. Fetal renal excretion of sodium was increased while urine volume decreased in the HSD group. Fetal blood urea nitrogen was increased, while kidney weight:body weight ratio decreased in the HSD group. The altered ratio was also observed in the offspring aged 15 and 90 days. Maternal and fetal plasma antidiuretic hormone was elevated without changes in plasma renin activity and Ang I levels, while plasma Ang II was decreased. The key elements of local renal RAS, including angiotensinogen, angiotensin converting enzyme (ACE), ACE2, AT1, and AT2 receptor expression in both mRNA and protein, except renin, were altered following maternal high salt intake. The results suggest that high intake of salt during pregnancy affected fetal renal development associated with an altered expression of the renal key elements of RAS, some alterations of fetal origins remained after birth as possible risks in developing renal or cardiovascular diseases.

  5. Assessment of a Salt Reduction Intervention on Adult Population Salt Intake in Fiji

    Directory of Open Access Journals (Sweden)

    Arti Pillay

    2017-12-01

    Full Text Available Reducing population salt intake is a global public health priority due to the potential to save lives and reduce the burden on the healthcare system through decreased blood pressure. This implementation science research project set out to measure salt consumption patterns and to assess the impact of a complex, multi-faceted intervention to reduce population salt intake in Fiji between 2012 and 2016. The intervention combined initiatives to engage food businesses to reduce salt in foods and meals with targeted consumer behavior change programs. There were 169 participants at baseline (response rate 28.2% and 272 at 20 months (response rate 22.4%. The mean salt intake from 24-h urine samples was estimated to be 11.7 grams per day (g/d at baseline and 10.3 g/d after 20 months (difference: −1.4 g/day, 95% CI −3.1 to 0.3, p = 0.115. Sub-analysis showed a statistically significant reduction in female salt intake in the Central Division but no differential impact in relation to age or ethnicity. Whilst the low response rate means it is not possible to draw firm conclusions about these changes, the population salt intake in Fiji, at 10.3 g/day, is still twice the World Health Organization’s (WHO recommended maximum intake. This project also assessed iodine intake levels in women of child-bearing age and found that they were within recommended guidelines. Existing policies and programs to reduce salt intake and prevent iodine deficiency need to be maintained or strengthened. Monitoring to assess changes in salt intake and to ensure that iodine levels remain adequate should be built into future surveys.

  6. MARS 1.3 system analysis code coupling with CONTEMPT4/MOD5/PCCS containment analysis code using dynamic link library

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Jeong, Jae Jun; Lee, Won Jae

    1998-01-01

    The two independent codes, MARS 1.3 and CONTEMPT4/MOD5/PCCS, have been coupled using the method of dynamic-link-library (DLL) technique. Overall configuration of the code system is designed so that MARS will be a main driver program which use CONTEMPT as associated routines. Using Digital Visual Fortran compiler, DLL was generated from the CONTEMPT source code with the interfacing routine names and arguments. Coupling of MARS with CONTEMPT was realized by calling the DLL routines at the appropriate step in the MARS code. Verification of coupling was carried out for LBLOCA transient of a typical plant design. It was found that the DLL technique is much more convenient than the UNIX process control techniques and effective for Window operating system. Since DLL can be used by more than one application and an application program can use many DLLs simultaneously, this technique would enable the existing codes to use more broadly with linking others

  7. MARS 1.3 system analysis code coupling with CONTEMPT4/MOD5/PCCS containment analysis code using dynamic link library

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Jeong, Jae Jun; Lee, Won Jae [KAERI, Taejon (Korea, Republic of)

    1998-10-01

    The two independent codes, MARS 1.3 and CONTEMPT4/MOD5/PCCS, have been coupled using the method of dynamic-link-library (DLL) technique. Overall configuration of the code system is designed so that MARS will be a main driver program which use CONTEMPT as associated routines. Using Digital Visual Fortran compiler, DLL was generated from the CONTEMPT source code with the interfacing routine names and arguments. Coupling of MARS with CONTEMPT was realized by calling the DLL routines at the appropriate step in the MARS code. Verification of coupling was carried out for LBLOCA transient of a typical plant design. It was found that the DLL technique is much more convenient than the UNIX process control techniques and effective for Window operating system. Since DLL can be used by more than one application and an application program can use many DLLs simultaneously, this technique would enable the existing codes to use more broadly with linking others.

  8. Coupling of 3D neutronics models with the system code ATHLET

    International Nuclear Information System (INIS)

    Langenbuch, S.; Velkov, K.

    1999-01-01

    The system code ATHLET for plant transient and accident analysis has been coupled with 3D neutronics models, like QUABOX/CUBBOX, for the realistic evaluation of some specific safety problems under discussion. The considerations for the coupling approach and its realization are discussed. The specific features of the coupled code system established are explained and experience from first applications is presented. (author)

  9. Field-based tests of geochemical modeling codes: New Zealand hydrothermal systems

    International Nuclear Information System (INIS)

    Bruton, C.J.; Glassley, W.E.; Bourcier, W.L.

    1993-12-01

    Hydrothermal systems in the Taupo Volcanic Zone, North Island, New Zealand are being used as field-based modeling exercises for the EQ3/6 geochemical modeling code package. Comparisons of the observed state and evolution of the hydrothermal systems with predictions of fluid-solid equilibria made using geochemical modeling codes will determine how the codes can be used to predict the chemical and mineralogical response of the environment to nuclear waste emplacement. Field-based exercises allow us to test the models on time scales unattainable in the laboratory. Preliminary predictions of mineral assemblages in equilibrium with fluids sampled from wells in the Wairakei and Kawerau geothermal field suggest that affinity-temperature diagrams must be used in conjunction with EQ6 to minimize the effect of uncertainties in thermodynamic and kinetic data on code predictions

  10. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    Energy Technology Data Exchange (ETDEWEB)

    Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-05-15

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  11. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    International Nuclear Information System (INIS)

    Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-01-01

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  12. Code division multiple-access techniques in optical fiber networks. II - Systems performance analysis

    Science.gov (United States)

    Salehi, Jawad A.; Brackett, Charles A.

    1989-08-01

    A technique based on optical orthogonal codes was presented by Salehi (1989) to establish a fiber-optic code-division multiple-access (FO-CDMA) communications system. The results are used to derive the bit error rate of the proposed FO-CDMA system as a function of data rate, code length, code weight, number of users, and receiver threshold. The performance characteristics for a variety of system parameters are discussed. A means of reducing the effective multiple-access interference signal by placing an optical hard-limiter at the front end of the desired optical correlator is presented. Performance calculations are shown for the FO-CDMA with an ideal optical hard-limiter, and it is shown that using a optical hard-limiter would, in general, improve system performance.

  13. Development of a domestically-made system code

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    According to lessons learned from the Fukushima-Daiichi NPP accidents, a new safety standard based on state-of-the-art findings has been established by the Japanese Nuclear Regulation Authority (NRA) and will soon come into force in Japan. In order to ensure a precise response to this movement from a technological point of view, it should be required for safety regulation to develop a new system code with much smaller uncertainty and reinforced simulation capability even in application to beyond-DBAs (BDBAs), as well as with the capability of close coupling to a newly developing severe accident code. Accordingly, development of a new domestically-made system code that incorporates 3-dimensional and 3 or more fluid thermal-hydraulics in tandem with a 3-dimensional neutronics has been started in 2012. In 2012, two branches of development activities, the development of 'main body' and advanced features have been started in parallel for development efficiency. The main body has been started from scratch and the following activities have therefore been performed: 1) development and determination of key principles and methodologies to realize a flexible, extensible and robust platform, 2) determination of requirements definition, 3) start of basic program design and coding and 4) start of a development of prototypical GUI-based pre-post processor. As for the advanced features, the following activities have been performed: 1) development of Phenomena Identification and Ranking Tables (PIRTs) and model capability matrix from normal operations to BDBAs in order to address requirements definition for advanced modeling, 2) development of detailed action plan for modification of field equations, numerical schemes and solvers and 3) start of the program development of field equations with an interfacial area concentration transport equation, a robust solver for condensation induced water hammer phenomena and a versatile Newton-Raphson solver. (author)

  14. Partitioning of L-methionine in aqueous two-phase systems containing poly(propylene glycol) and sodium phosphate salts

    International Nuclear Information System (INIS)

    Salabat, Alireza; Sadeghi, Rahmat; Moghadam, Somayeh Tiani; Jamehbozorg, Bahman

    2011-01-01

    Highlights: → Thermodynamics parameters for partitioning of L-methionine in ATPS. → Investigation of different effects on partition coefficient of the amino acid. → Propose the best condition for L-methionine partitioning. - Abstract: The partitioning behavior of L-methionine has been studied in aqueous two-phase systems of (poly(propylene glycol) + sodium phosphate salts + H 2 O) at different temperatures. The salts used were sodium di-hydrogen phosphate (NaH 2 PO 4 ), di-sodium hydrogen phosphate (Na 2 HPO 4 ) and tri-sodium phosphate (Na 3 PO 4 ). The effects of tie line length, salt type, and temperature on the partition coefficient of this amino acid have been studied. In addition, thermodynamic parameters (ΔH o , ΔS o and ΔG o ) as a function of temperature were calculated. The results showed that increasing tie line length led to decreasing of the partition coefficient. We also showed that the partition coefficients of the amino acid in the systems containing Na 3 PO 4 are greater than the other two salts. Moreover, it is verified that increasing temperature led to decreasing the partition coefficient. The experimental partition coefficient data are correlated using a modified virial-type model.

  15. Rural Latino caregivers' beliefs and behaviors around their children's salt consumption.

    Science.gov (United States)

    Hoeft, Kristin S; Guerra, Claudia; Gonzalez-Vargas, M Judy; Barker, Judith C

    2015-04-01

    Prevalence of high blood pressure has been increasing in U.S. children, with implications for long term health consequences. Sodium consumption, a modifiable risk factor for high blood pressure, is above recommended limits and increasing. Very little is known about Latino caregiver beliefs and behaviors around their children's salt consumption. In California's Central Valley, qualitative interviews in Spanish investigated low-income caregivers' views and understandings of their children's dietary salt consumption. Thirty individual interviews and 5 focus groups were conducted (N=61). Interview transcripts were translated and transcribed, coded and thematically analyzed. Seven primary topic areas around children's salt intake and its impact on health were identified: children's favorite foods, children's dietary salt sources, superiority of home-cooked foods, salty and sweet foods, managing salt for health, developing children's tastes, and adding salt added at the table. Parents recognize common sources of sodium such as "junk food" and processed food and made efforts to limit their children's consumption of these foods, but may overlook other significant sodium sources, particularly bread, cheese, prepared soups and sports drinks. Caregivers recognize excess salt as unhealthy for children, but don't believe health problems (like high blood pressure) can occur in young children. Nevertheless, they made efforts to limit how much salt their children consumed through a variety of strategies; school meals were a source of high sodium that they felt were outside of their control. Latino caregivers are concerned about their children's salt intake and attempt to limit consumption, but some common sources of sodium are under-recognized. Copyright © 2014 Elsevier Ltd. All rights reserved.

  16. Molten salt engineering for thorium cycle. Electrochemical studies as examples

    International Nuclear Information System (INIS)

    Ito, Yasuhiko

    1998-01-01

    A Th-U nuclear energy system utilizing accelerator driven subcritical molten salt breeder reactor has several advantages compared to conventional U-Pu nuclear system. In order to obtain fundamental data on molten salt engineering of Th-U system, electrochemical study was conducted. As the most primitive simulated study of beam irradiation of molten salt, discharge electrolysis was investigated in molten LiCl-KCl-AgCl system. Stationary discharge was generated under atmospheric argon gas and fine Ag particles were obtained. Hydride ion (H - ) behavior in molten salts was also studied to predict the behavior of tritide ion (T - ) in molten salt fuel. Finally, hydrogen behavior in metals at high temperature was investigated by electrochemical method, which is considered to be important to confine and control tritium. (author)

  17. Molten salts in nuclear reactors

    International Nuclear Information System (INIS)

    Dirian, J.; Saint-James

    1959-01-01

    Collection of references dealing with the physicochemical studies of fused salts, in particular the alkali and alkali earth halides. Numerous binary, ternary and quaternary systems of these halides with those of uranium and thorium are examined, and the physical properties, density, viscosity, vapour pressure etc... going from the halides to the mixtures are also considered. References relating to the corrosion of materials by these salts are included and the treatment of the salts with a view to recuperation after irradiation in a nuclear reactor is discussed. (author) [fr

  18. Comparison of an antioxidant system in tolerant and susceptible wheat seedlings in response to salt stress

    Energy Technology Data Exchange (ETDEWEB)

    Feki, K.; Tounsi, S.; Brini, F.

    2017-07-01

    It has been demonstrated previously that the physiological and molecular analysis of seedlings of the tolerant (Om Rabia3) and susceptible (Mahmoudi) Tunisian wheat genotypes were different at short and long-term response to salinity. In this study, we examined the antioxidant defence system in seedlings of these two cultivars at short-term response to different NaCl concentrations. The findings showed that high salinity tolerance of cv. Om Rabia3, as manifested by lower decrease in its dry biomass, was associated with lower malondialdehyde and hydrogen peroxide contents, lower accumulation of the superoxide (O2⎯) in the roots and the shoots, and also lower decrease in ascorbate content than those in cv. Mahmoudi. Moreover, the expression of some genes coding for antioxidant enzymes such as the catalase, the superoxide dismutase and the peroxidase were enhanced by NaCl stress especially in the salt-tolerant cultivar. In parallel, their activities were increased in response to the same condition of stress and especially in the cv. Om Rabia3. Taken together, these data suggested that the capacity to limit oxidative damage is important for NaCl tolerance of durum wheat.

  19. Comparison of the genetic organization of the early salt-stress-response gene system in salt-tolerant Lophopyrum elongatum and salt-sensitive wheat

    OpenAIRE

    Dubcovsky, J; Galvez, AF; Dvořák, J

    1994-01-01

    Lophopyrum elongatum is a facultative halophyte related to wheat. Eleven unique clones corresponding to genes showing enhanced mRNA accumulation in the early stages of salt stress were previously isolated from a L. elongatum salt-stressed-root cDNA library. The chromosomal distribution of genes complementary to these clones in several genomes of the tribe Triticeae and their copy number in the L. elongatum and wheat genomes are reported. Genes complementary to clones pESI4, pESI14, pESI15, pE...

  20. Analytical considerations in the code qualification of piping systems

    International Nuclear Information System (INIS)

    Antaki, G.A.

    1995-01-01

    The paper addresses several analytical topics in the design and qualification of piping systems which have a direct bearing on the prediction of stresses in the pipe and hence on the application of the equations of NB, NC and ND-3600 of the ASME Boiler and Pressure Vessel Code. For each of the analytical topics, the paper summarizes the current code requirements, if any, and the industry practice

  1. JPEG2000 COMPRESSION CODING USING HUMAN VISUAL SYSTEM MODEL

    Institute of Scientific and Technical Information of China (English)

    Xiao Jiang; Wu Chengke

    2005-01-01

    In order to apply the Human Visual System (HVS) model to JPEG2000 standard,several implementation alternatives are discussed and a new scheme of visual optimization isintroduced with modifying the slope of rate-distortion. The novelty is that the method of visual weighting is not lifting the coefficients in wavelet domain, but is complemented by code stream organization. It remains all the features of Embedded Block Coding with Optimized Truncation (EBCOT) such as resolution progressive, good robust for error bit spread and compatibility of lossless compression. Well performed than other methods, it keeps the shortest standard codestream and decompression time and owns the ability of VIsual Progressive (VIP) coding.

  2. Rotated Walsh-Hadamard Spreading with Robust Channel Estimation for a Coded MC-CDMA System

    Directory of Open Access Journals (Sweden)

    Raulefs Ronald

    2004-01-01

    Full Text Available We investigate rotated Walsh-Hadamard spreading matrices for a broadband MC-CDMA system with robust channel estimation in the synchronous downlink. The similarities between rotated spreading and signal space diversity are outlined. In a multiuser MC-CDMA system, possible performance improvements are based on the chosen detector, the channel code, and its Hamming distance. By applying rotated spreading in comparison to a standard Walsh-Hadamard spreading code, a higher throughput can be achieved. As combining the channel code and the spreading code forms a concatenated code, the overall minimum Hamming distance of the concatenated code increases. This asymptotically results in an improvement of the bit error rate for high signal-to-noise ratio. Higher convolutional channel code rates are mostly generated by puncturing good low-rate channel codes. The overall Hamming distance decreases significantly for the punctured channel codes. Higher channel code rates are favorable for MC-CDMA, as MC-CDMA utilizes diversity more efficiently compared to pure OFDMA. The application of rotated spreading in an MC-CDMA system allows exploiting diversity even further. We demonstrate that the rotated spreading gain is still present for a robust pilot-aided channel estimator. In a well-designed system, rotated spreading extends the performance by using a maximum likelihood detector with robust channel estimation at the receiver by about 1 dB.

  3. Structure and operation of the ITS code system

    International Nuclear Information System (INIS)

    Halbleib, J.

    1988-01-01

    The TIGER series of time-independent coupled electron-photon Monte Carlo transport codes is a group of multimaterial and multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron-photon cascade by combining microscopic photon transport with a macroscopic random walk for electron transport. Major contributors to its evolution are listed. The author and his associates are primarily code users rather than code developers, and have borrowed freely from existing work wherever possible. Nevertheless, their efforts have resulted in various software packages for describing the production and transport of the electron-photon cascade that they found sufficiently useful to warrant dissemination through the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory. The ITS system (Integrated TIGER Series) represents the organization and integration of this combined software, along with much additional capability from previously unreleased work, into a single convenient package of exceptional user friendliness and portability. Emphasis is on simplicity and flexibility of application without sacrificing the rigor or sophistication of the physical model

  4. A seismic data compression system using subband coding

    Science.gov (United States)

    Kiely, A. B.; Pollara, F.

    1995-01-01

    This article presents a study of seismic data compression techniques and a compression algorithm based on subband coding. The algorithm includes three stages: a decorrelation stage, a quantization stage that introduces a controlled amount of distortion to allow for high compression ratios, and a lossless entropy coding stage based on a simple but efficient arithmetic coding method. Subband coding methods are particularly suited to the decorrelation of nonstationary processes such as seismic events. Adaptivity to the nonstationary behavior of the waveform is achieved by dividing the data into separate blocks that are encoded separately with an adaptive arithmetic encoder. This is done with high efficiency due to the low overhead introduced by the arithmetic encoder in specifying its parameters. The technique could be used as a progressive transmission system, where successive refinements of the data can be requested by the user. This allows seismologists to first examine a coarse version of waveforms with minimal usage of the channel and then decide where refinements are required. Rate-distortion performance results are presented and comparisons are made with two block transform methods.

  5. The electrodeposition and rare earths reduction in the molten salt actinides recovery systems using liquid metal

    International Nuclear Information System (INIS)

    Shim, J-B.; Lee, J-H.; Kwon, S-W.; Ahn, B-G.; Woo, M-S.; Lee, B-J.; Kim, E-H.; Park, H-S.; Yoo, J-H.

    2005-01-01

    A pyrochemical partitioning system uses liquid metals such as cadmium and bismuth in order to recover the actinide metals from a molten salt mixture containing rare earth fission product metals. The liquid metals play roles as a cathode in the electrowinning or an extracting phase in the reductive extraction operation. The product resulting from the above operations is metal-cadmium or-bismuth alloy, which should contain the rare earth element amounts as low as possible for a transmutation purpose. In this study, the electrodeposition behaviours of uranium and lanthanide elements such as La, Ce and Nd were investigated for solid molybdenum and liquid cadmium electrodes in a molten LiCl-KCl eutectic salt. Electrochemical methods used are a cyclic voltammetry (CV) and a chronopotentiometry for monitoring the salt phase and recovering the metals, respectively. The CV graphs for monitoring the oxidizing agent CdCl 2 in the salt phase were obtained. These show a time dependently disappearance of the oxidizing agent corresponding to the formation of UCl 3 by inserting the uranium metal into the salt. Also, a sequential oxidation technique which is added at a controlled amount of the oxidizing agents into the salt phase was applied. It was found that this method is feasible for the selective reduction of the rare earths content in liquid metal alloys. (author)

  6. Systematic Evaluation of Salt Cavern Well Integrity

    Science.gov (United States)

    Roberts, B. L.; Lord, D. L.; Lord, A. S.; Bettin, G.; Sobolik, S. R.; Park, B. Y.

    2017-12-01

    The U.S. Strategic Petroleum Reserve (SPR) holds a reserve of crude oil ( 700 million barrels) to help ease any interruptions in oil import to the United States. The oil is stored in a set of 63 underground caverns distributed across four sites along the U.S. Gulf Coast. The caverns were solution mined into salt domes at each of the four sites. The plastic nature of the salt is beneficial for the storage of crude oil as it heals any fractures that may occur in the salt. The SPR is responsible for operating and maintaining the nearly 120 wells used to access the storage caverns over operational lifetimes spanning decades. Salt creep can induce deformation of the well casing which must be remediated to insure cavern and well integrity. This is particularly true at the interface between the plastic salt and the rigid caprock. The Department of Energy, the SPR Management and Operations contractor, and Sandia National Laboratories has developed a multidimensional well-grading system for the salt cavern access wells. This system is designed to assign numeric grades to each well indicating its risk of losing integrity and remediation priority. The system consists of several main components which themselves may consist of sub-components. The main components consider such things as salt cavern pressure history, results from geomechanical simulations modeling salt deformation, and measurements of well casing deformation due to salt creep. In addition, the geology of the salt domes and their overlying caprock is also included in the grading. These multiple factors are combined into summary values giving the monitoring and remediation priority for each well. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia LLC, a wholly owned subsidiary of Honeywell International Inc. for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-NA0003525.

  7. Design study on advanced nuclear fuel recycle system. Conceptual design study of recycle system using molten salt

    International Nuclear Information System (INIS)

    Kasai, Y.; Kakehi, I.; Moro, T.; Higashi, T.; Tobe, K.; Kawamura, F.; Yonezawa, S.; Yoshiuji, T.

    1998-10-01

    Advanced recycle system engineering group of OEC (Oarai Engineering Center) has being carried out a design study of the advanced nuclear fuel recycle system using molten salt (electro-metallurgical process). This system is aiming for improvements of fuel cycle economy and reduction of environmental burden (MA recycles, Minimum of radioactive waste disposal), and also improvement of safety and nuclear non-proliferation. This report describes results of the design study that has been continued since December 1996. (1) A design concept of the advanced nuclear fuel recycle system, that is a module type recycles system of pyrochemical reprocessing and fuel re-fabrication was studied. The module system has advantage in balance of Pu recycle where modules are constructed in coincidence with the construction plan of nuclear power plants, and also has flexibility for technology progress. A demonstration system, minimum size of the above module, was studies. This system has capacity of 10 tHM/y and is able to demonstrate recycle technology of MOX fuel, metal fuel and nitride fuel. (2) Each process of the system, which are pyrochemical electrorefining system, cathode processor, de-cladding system, waste disposal system, etc., were studied. In this study, capacity of an electrorefiner was discussed, and vitrification experiment of molten salt using lead-boric acid glass was conducted. (3) A hot cell system and material handling system of the demonstration system was studied. A robot driven by linear motor was studied for the handling system, and an arrangement plan of the cell system was made. Criticality analysis in the cell system and investigation of material accountancy system of the recycle plant were also made. This design study will be continued in coincidence with design study of reactor and fuel, aiming to establish the concept of FBR recycle system. (author)

  8. Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my [Centre of Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Cioncolini, Andrea; Iacovides, Hector [School of Mechanical, Aerospace, and Civil Engineering (MACE), University of Manchester, Oxford Road, M13 9PL Manchester (United Kingdom)

    2015-04-29

    The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

  9. Mass transport in bedded salt and salt interbeds

    International Nuclear Information System (INIS)

    Hwang, Y.; Pigford, T.H.; Chambre, P.L.; Lee, W.W.L.

    1989-08-01

    Salt is the proposed host rock for geologic repositories of nuclear waste in several nations because it is nearly dry and probably impermeable. Although experiments and experience at potential salt sites indicate that salt may contain brine, the low porosity, creep, and permeability of salt make it still a good choice for geologic isolation. In this paper we summarize several mass-transfer and transport analyses of salt repositories. The mathematical details are given in our technical reports

  10. LDPC concatenated space-time block coded system in multipath fading environment: Analysis and evaluation

    Directory of Open Access Journals (Sweden)

    Surbhi Sharma

    2011-06-01

    Full Text Available Irregular low-density parity-check (LDPC codes have been found to show exceptionally good performance for single antenna systems over a wide class of channels. In this paper, the performance of LDPC codes with multiple antenna systems is investigated in flat Rayleigh and Rician fading channels for different modulation schemes. The focus of attention is mainly on the concatenation of irregular LDPC codes with complex orthogonal space-time codes. Iterative decoding is carried out with a density evolution method that sets a threshold above which the code performs well. For the proposed concatenated system, the simulation results show that the QAM technique achieves a higher coding gain of 8.8 dB and 3.2 dB over the QPSK technique in Rician (LOS and Rayleigh (NLOS faded environments respectively.

  11. Design and implementation of a scene-dependent dynamically selfadaptable wavefront coding imaging system

    Science.gov (United States)

    Carles, Guillem; Ferran, Carme; Carnicer, Artur; Bosch, Salvador

    2012-01-01

    A computational imaging system based on wavefront coding is presented. Wavefront coding provides an extension of the depth-of-field at the expense of a slight reduction of image quality. This trade-off results from the amount of coding used. By using spatial light modulators, a flexible coding is achieved which permits it to be increased or decreased as needed. In this paper a computational method is proposed for evaluating the output of a wavefront coding imaging system equipped with a spatial light modulator, with the aim of thus making it possible to implement the most suitable coding strength for a given scene. This is achieved in an unsupervised manner, thus the whole system acts as a dynamically selfadaptable imaging system. The program presented here controls the spatial light modulator and the camera, and also processes the images in a synchronised way in order to implement the dynamic system in real time. A prototype of the system was implemented in the laboratory and illustrative examples of the performance are reported in this paper. Program summaryProgram title: DynWFC (Dynamic WaveFront Coding) Catalogue identifier: AEKC_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEKC_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 10 483 No. of bytes in distributed program, including test data, etc.: 2 437 713 Distribution format: tar.gz Programming language: Labview 8.5 and NI Vision and MinGW C Compiler Computer: Tested on PC Intel ® Pentium ® Operating system: Tested on Windows XP Classification: 18 Nature of problem: The program implements an enhanced wavefront coding imaging system able to adapt the degree of coding to the requirements of a specific scene. The program controls the acquisition by a camera, the display of a spatial light modulator

  12. [How can the coding quality in the DRG system be determined?].

    Science.gov (United States)

    Kahlmeyer, A; Volkmer, B

    2014-01-01

    The permanent adjustments ​​since 2003 to the G-DRG system have made the system even less understandable, so that many users have the feeling of feeding data into a black box which gives them a result without them being able to actively use the system itself. While chief physicians, senior physicians, and nursing managers are responsible to management for the results of the billing, they are in most cases not involved in the steps of DRG coding and billing. From this situation, a common question arises: "How well does my department code?" This uncertainty is exploited by many commercial vendors, who offer a wide variety of approaches for DRG optimization. The goal of this work is to provide advice as to how coding quality can be determined.

  13. Salt disposal of heat-generating nuclear waste

    International Nuclear Information System (INIS)

    Leigh, Christi D.; Hansen, Francis D.

    2011-01-01

    This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United

  14. Salt disposal of heat-generating nuclear waste.

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, Christi D. (Sandia National Laboratories, Carlsbad, NM); Hansen, Francis D.

    2011-01-01

    This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from

  15. THYDE-NEU: Nuclear reactor system analysis code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2002-03-01

    THYDE-NEU is applicable not only to transient analyses, but also to steady state analyses of nuclear reactor systems (NRSs). In a steady state analysis, the code generates a solution satisfying the transient equations without external disturbances. In a transient analysis, the code calculates temporal NRS behaviors in response to various external disturbances in such a way that mass and energy of the coolant as well as the number of neutrons conserve. The first half of the report is the description of the methods and models for use in the THYDE-NEU code, i.e., (1) the thermal-hydraulic network model, (2) the spatial kinetics model, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the users' mannual containing the items; (1) the program control, (2) the input requirements, (3) the execution of THYDE-NEU jobs, (4) the output specifications and (5) the sample calculation. (author)

  16. Waste treatment using molten salt oxidation

    International Nuclear Information System (INIS)

    Navratil, J.D.; Stewart, A.E.

    1996-01-01

    MSO technology can be characterized as a submerged oxidation process; the basic concept is to introduce air and wastes into a bed of molten salt, oxidize the organic wastes in the molten salt, use the heat of oxidation to keep the salt molten and remove the salt for disposal or processing and recycling. The molten salt (usually sodium carbonate at 900-1000 C) provides four waste management functions: providing a heat transfer medium, catalyzing the oxidation reaction, preventing the formation of acid gases by forming stable salts, and efficiently capturing ash particles and radioactive materials by the combined effects of wetting, encapsulation and dissolution. The MSO process requires no wet scrubbing system for off-gas treatment. The process has been developed through bench-scale and pilot-scale testing, with successful destruction demonstration of a wide variety of hazardous and mixed (radioactive and hazardous wastes). (author). 24 refs, 2 tabs, 2 figs

  17. Two- and Three-Component Visible Light Photoinitiating Systems for Radical Polymerization Based on Onium Salts: An Overview of Mechanistic and Laser Flash Photolysis Studies

    Directory of Open Access Journals (Sweden)

    María L. Gómez

    2012-01-01

    Full Text Available A review of our work on two- and-three component photoinitiator systems is presented. The emphasis is in on visible light polymerization in aqueous media. The systems discussed comprise a synthetic dye as sensitizer and an onium salt as coinitiator, or a dye-amine-onium salt with the amine as coinitiator and the onium salt as an enhancer of the polymerization efficiency. The effect of the composition of the system on the photopolymerization kinetics was analyzed. To this end, the photophysics and photochemistry of the dye under polymerization conditions was explored by means of stationary and time-resolved spectroscopic methods. Different dyes and onium salts were investigated. The action mechanism of the different photoinitiators systems is discussed.

  18. Simulation realization of 2-D wavelength/time system utilizing MDW code for OCDMA system

    Science.gov (United States)

    Azura, M. S. A.; Rashidi, C. B. M.; Aljunid, S. A.; Endut, R.; Ali, N.

    2017-11-01

    This paper presents a realization of Wavelength/Time (W/T) Two-Dimensional Modified Double Weight (2-D MDW) code for Optical Code Division Multiple Access (OCDMA) system based on Spectral Amplitude Coding (SAC) approach. The MDW code has the capability to suppress Phase-Induce Intensity Noise (PIIN) and minimizing the Multiple Access Interference (MAI) noises. At the permissible BER 10-9, the 2-D MDW (APD) had shown minimum effective received power (Psr) = -71 dBm that can be obtained at the receiver side as compared to 2-D MDW (PIN) only received -61 dBm. The results show that 2-D MDW (APD) has better performance in achieving same BER with longer optical fiber length and with less received power (Psr). Also, the BER from the result shows that MDW code has the capability to suppress PIIN ad MAI.

  19. Simulation realization of 2-D wavelength/time system utilizing MDW code for OCDMA system

    Directory of Open Access Journals (Sweden)

    Azura M. S. A.

    2017-01-01

    Full Text Available This paper presents a realization of Wavelength/Time (W/T Two-Dimensional Modified Double Weight (2-D MDW code for Optical Code Division Multiple Access (OCDMA system based on Spectral Amplitude Coding (SAC approach. The MDW code has the capability to suppress Phase-Induce Intensity Noise (PIIN and minimizing the Multiple Access Interference (MAI noises. At the permissible BER 10-9, the 2-D MDW (APD had shown minimum effective received power (Psr = -71 dBm that can be obtained at the receiver side as compared to 2-D MDW (PIN only received -61 dBm. The results show that 2-D MDW (APD has better performance in achieving same BER with longer optical fiber length and with less received power (Psr. Also, the BER from the result shows that MDW code has the capability to suppress PIIN ad MAI.

  20. Separate Turbo Code and Single Turbo Code Adaptive OFDM Transmissions

    Directory of Open Access Journals (Sweden)

    Burr Alister

    2009-01-01

    Full Text Available Abstract This paper discusses the application of adaptive modulation and adaptive rate turbo coding to orthogonal frequency-division multiplexing (OFDM, to increase throughput on the time and frequency selective channel. The adaptive turbo code scheme is based on a subband adaptive method, and compares two adaptive systems: a conventional approach where a separate turbo code is used for each subband, and a single turbo code adaptive system which uses a single turbo code over all subbands. Five modulation schemes (BPSK, QPSK, 8AMPM, 16QAM, and 64QAM are employed and turbo code rates considered are and . The performances of both systems with high ( and low ( BER targets are compared. Simulation results for throughput and BER show that the single turbo code adaptive system provides a significant improvement.

  1. Ultrasonic Technique for Predicting Grittiness of Salted Duck Egg

    Science.gov (United States)

    Erawan, S.; Budiastra, I. W.; Subrata, I. D. M.

    2018-05-01

    Grittiness of egg yolk is a major factor in consumer acceptance of salted duck egg product. Commonly, the grittiness level is determined by the destructive method. Salted egg industries need a grading system that can judge the grittiness accurately and nondestructively. The purpose of this study was to develop a method for determining grittiness of salted duck eggs nondestructively based on ultrasonic method. This study used 100 samples of salted duck eggs with 7,10,14 and 21 days of salting age. Velocity and attenuation were measured by an ultrasonic system at frequency 50 kHz, followed by physicochemical properties measurement (hardness of egg yolks and salt content), and organoleptic test. Ultrasonic wave velocity in salted duck eggs ranged from 620.6 m/s to 1334.6 m/s, while the coefficient of attenuation value ranged from – 0.76 dB/m to -0.51 dB/m. Yolk hardness was 2.68 N at 7 days to 5.54 N at 21 days of salting age. Salt content was 1.81 % at 7 days to 5.71 % at 21 days of salting age. Highest scores of organoleptic tests on salted duck eggs were 4.23 and 4.18 for 10 and 14 days of salting age, respectively. Discriminant function using ultrasonic velocity variables in minor and major diameter could predict grittiness with 95 % accuracy.

  2. Large-scale dynamic compaction of natural salt

    International Nuclear Information System (INIS)

    Hansen, F.D.; Ahrens, E.H.

    1996-01-01

    A large-scale dynamic compaction demonstration of natural salt was successfully completed. About 40 m 3 of salt were compacted in three, 2-m lifts by dropping a 9,000-kg weight from a height of 15 m in a systematic pattern to achieve desired compaction energy. To enhance compaction, 1 wt% water was added to the relatively dry mine-run salt. The average compacted mass fractional density was 0.90 of natural intact salt, and in situ nitrogen permeabilities averaged 9X10 -14 m 2 . This established viability of dynamic compacting for placing salt shaft seal components. The demonstration also provided compacted salt parameters needed for shaft seal system design and performance assessments of the Waste Isolation Pilot Plant

  3. Supra-molecular Association and Polymorphic Behaviour In Systems Containing Bile Acid Salts

    Directory of Open Access Journals (Sweden)

    Camillo La Mesa

    2007-08-01

    Full Text Available A wide number of supra-molecular association modes are observed in mixtures containing water and bile salts, BS, (with, eventually, other components. Molecular or micellar solutions transform into hydrated solids, fibres, lyotropic liquid crystals and/or gels by raising the concentration, the temperature, adding electrolytes, surfactants, lipids and proteins. Amorphous or ordered phases may be formed accordingly. The forces responsible for this very rich polymorphism presumably arise from the unusual combination of electrostatic, hydrophobic and hydrogen-bond contributions to the system stability, with subsequent control of the supra-molecular organisation modes. The stabilising effect due to hydrogen bonds does not occur in almost all surfactants or lipids and is peculiar to bile acids and salts. Some supra-molecular organisation modes, supposed to be related to malfunctions and dis-metabolic diseases in vivo, are briefly reported and discussed.

  4. Management of manufacture and installation of plant pipings by bar code system

    International Nuclear Information System (INIS)

    Suwa, Minoru

    1995-01-01

    As for the piping system of nuclear power plants, the number of parts is very large, and the mill sheet is attached to each part, therefore, it is necessary to manage them individually, and large man power is required. In order to resolve the delay of mechanization in the factory, bar code system was adopted on full scale. At the time of taking parts out from the store, bar code labels are stuck to all piping parts. By this means, all the processes of manufacture and inspection are managed with a computer, and it is useful for labor saving and the prevention of mistaken input. This system is centering around the system of the progress management for piping manufacture, and is operated by being coupled with respective systems of production design, order and inventory, mill sheet management and installation management. The management of production design, manufacture, inspection and installation is explained. There is the problem of sticking bar code labels again as the labels become dirty or parts pass through coating and pickling processes. The direct carving of bar codes on parts by laser marker was tried, and it was successful for stainless steel, but in carbon steel pipes, it was hard to read. It is desirable to develop the bar codes which endure until the end of plant life. (K.I.)

  5. Extraction mechanism of sulfamethoxazole in water samples using aqueous two-phase systems of poly(propylene glycol) and salt

    Energy Technology Data Exchange (ETDEWEB)

    Xie Xueqiao; Wang Yun; Han Juan [School of Chemistry and Chemical Engineering, Jiangsu University, Zhenjiang 212013 (China); Yan Yongsheng, E-mail: yys@ujs.edu.cn [School of Chemistry and Chemical Engineering, Jiangsu University, Zhenjiang 212013 (China)

    2011-02-14

    Based on the poly(propylene glycol){sub 400} (PPG{sub 400})-salt aqueous two-phase system (ATPS), a green, economical and effective sample pretreatment technique coupled with high performance liquid chromatography was proposed for the separation and determination of sulfamethoxazole (SMX). The extraction yield of SMX in PPG{sub 400}-salt ATPS is influenced by various factors, including the salt species, the amount of salt, pH, and the temperature. Under the optimum conditions, most of SMX was partitioning into the polymer-rich phase with the average extraction efficiency of 99.2%, which may be attributed to the hydrophobic interaction and salting-out effect. This extraction technique has been successfully applied to the analysis of SMX in real water samples with the recoveries of 96.0-100.6%, the detection limits of 0.1 {mu}g L{sup -1}, and the linear ranges of 2.5-250.0 {mu}g L{sup -1}.

  6. PERFORMANCE ANALYSIS OF OPTICAL CDMA SYSTEM USING VC CODE FAMILY UNDER VARIOUS OPTICAL PARAMETERS

    Directory of Open Access Journals (Sweden)

    HASSAN YOUSIF AHMED

    2012-06-01

    Full Text Available The intent of this paper is to study the performance of spectral-amplitude coding optical code-division multiple-access (OCDMA systems using Vector Combinatorial (VC code under various optical parameters. This code can be constructed by an algebraic way based on Euclidian vectors for any positive integer number. One of the important properties of this code is that the maximum cross-correlation is always one which means that multi-user interference (MUI and phase induced intensity noise are reduced. Transmitter and receiver structures based on unchirped fiber Bragg grating (FBGs using VC code and taking into account effects of the intensity, shot and thermal noise sources is demonstrated. The impact of the fiber distance effects on bit error rate (BER is reported using a commercial optical systems simulator, virtual photonic instrument, VPITM. The VC code is compared mathematically with reported codes which use similar techniques. We analyzed and characterized the fiber link, received power, BER and channel spacing. The performance and optimization of VC code in SAC-OCDMA system is reported. By comparing the theoretical and simulation results taken from VPITM, we have demonstrated that, for a high number of users, even if data rate is higher, the effective power source is adequate when the VC is used. Also it is found that as the channel spacing width goes from very narrow to wider, the BER decreases, best performance occurs at a spacing bandwidth between 0.8 and 1 nm. We have shown that the SAC system utilizing VC code significantly improves the performance compared with the reported codes.

  7. Establishment of computer code system for nuclear reactor design - analysis

    International Nuclear Information System (INIS)

    Subki, I.R.; Santoso, B.; Syaukat, A.; Lee, S.M.

    1996-01-01

    Establishment of computer code system for nuclear reactor design analysis is given in this paper. This establishment is an effort to provide the capability in running various codes from nuclear data to reactor design and promote the capability for nuclear reactor design analysis particularly from neutronics and safety points. This establishment is also an effort to enhance the coordination of nuclear codes application and development existing in various research centre in Indonesia. Very prospective results have been obtained with the help of IAEA technical assistance. (author). 6 refs, 1 fig., 1 tab

  8. System transient analysis code development for low pressure and low power

    International Nuclear Information System (INIS)

    Kim, Hee Cheol

    1998-02-01

    A real time reactor system analysis code, ARTIST, based on drift flux model has been developed to investigate the transient system behavior under low pressure, low flow and low power conditions with noncondensable gas present in the system. The governing equations of the ARTIST code consist of three mass continuity equations (steam, liquid and noncondensable), two energy equations (gas and mixture) and one momentum equation (mixture) constituted with the drift flux model. The capability of ARTIST in predicting two-phase flow void distribution in the system has been validated against experimental data. The results of the ARTIST axial void distribution at low pressure and low flow, are far better than the results of both the homogeneous model of TASS code and the two-fluid model of RELAP5/MOD3 code. Also, RELAP5/MOD3 calculation shows the large amplitude of void fraction oscillations at low pressure. These results imply that interfacial momentum transfer terms in the two-fluid model formulation should be carefully constituted, especially for the low pressure condition due to the big density differences between steam and water. Thermal-hydraulic state solution scheme is developed when noncondensable gas exists. Numerical consistency and convergence of obtaining equilibrium state is tested with the ideal problems for various situations including very low partial pressure conditions. Calculated thermal-hydraulic state for each test shows consistent and expected behaviour. A new multi-layer back propagation network algorithm for calculating the departure from nucleate boiling ratio (DNBR) is developed and adopted in ARTIST code in order to have real-time DNBR evaluation by eliminating the tandem procedure of the transient DNBR calculation. The algorithm trained by different patterns generated by latin hypercube sampling method on the performance space is tested for the randomly sampled untrained data and the transient DNBR data. The uncertainty of the algorithm is

  9. A computerized energy systems code and information library at Soreq

    Energy Technology Data Exchange (ETDEWEB)

    Silverman, I; Shapira, M; Caner, D; Sapier, D [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center

    1996-12-01

    In the framework of the contractual agreement between the Ministry of Energy and Infrastructure and the Division of Nuclear Engineering of the Israel Atomic Energy Commission, both Soreq-NRC and Ben-Gurion University have agreed to establish, in 1991, a code center. This code center contains a library of computer codes and relevant data, with particular emphasis on nuclear power plant research and development support. The code center maintains existing computer codes and adapts them to the ever changing computing environment, keeps track of new code developments in the field of nuclear engineering, and acquires the most recent revisions of computer codes of interest. An attempt is made to collect relevant codes developed in Israel and to assure that proper documentation and application instructions are available. En addition to computer programs, the code center collects sample problems and international benchmarks to verify the codes and their applications to various areas of interest to nuclear power plant engineering and safety evaluation. Recently, the reactor simulation group at Soreq acquired, using funds provided by the Ministry of Energy and Infrastructure, a PC work station operating under a Linux operating system to give users of the library an easy on-line way to access resources available at the library. These resources include the computer codes and their documentation, reports published by the reactor simulation group, and other information databases available at Soreq. Registered users set a communication line, through a modem, between their computer and the new workstation at Soreq and use it to download codes and/or information or to solve their problems, using codes from the library, on the computer at Soreq (authors).

  10. A computerized energy systems code and information library at Soreq

    International Nuclear Information System (INIS)

    Silverman, I.; Shapira, M.; Caner, D.; Sapier, D.

    1996-01-01

    In the framework of the contractual agreement between the Ministry of Energy and Infrastructure and the Division of Nuclear Engineering of the Israel Atomic Energy Commission, both Soreq-NRC and Ben-Gurion University have agreed to establish, in 1991, a code center. This code center contains a library of computer codes and relevant data, with particular emphasis on nuclear power plant research and development support. The code center maintains existing computer codes and adapts them to the ever changing computing environment, keeps track of new code developments in the field of nuclear engineering, and acquires the most recent revisions of computer codes of interest. An attempt is made to collect relevant codes developed in Israel and to assure that proper documentation and application instructions are available. En addition to computer programs, the code center collects sample problems and international benchmarks to verify the codes and their applications to various areas of interest to nuclear power plant engineering and safety evaluation. Recently, the reactor simulation group at Soreq acquired, using funds provided by the Ministry of Energy and Infrastructure, a PC work station operating under a Linux operating system to give users of the library an easy on-line way to access resources available at the library. These resources include the computer codes and their documentation, reports published by the reactor simulation group, and other information databases available at Soreq. Registered users set a communication line, through a modem, between their computer and the new workstation at Soreq and use it to download codes and/or information or to solve their problems, using codes from the library, on the computer at Soreq (authors)

  11. Error-correction coding

    Science.gov (United States)

    Hinds, Erold W. (Principal Investigator)

    1996-01-01

    This report describes the progress made towards the completion of a specific task on error-correcting coding. The proposed research consisted of investigating the use of modulation block codes as the inner code of a concatenated coding system in order to improve the overall space link communications performance. The study proposed to identify and analyze candidate codes that will complement the performance of the overall coding system which uses the interleaved RS (255,223) code as the outer code.

  12. Thorium-based Molten Salt Reactor (TMSR) project in China

    International Nuclear Information System (INIS)

    Dai, Zhimin; Liu, Wei

    2013-01-01

    Making great efforts in development of nuclear energy is one of the long-term-plan in China's energy strategies. The advantages of Thorium-based nuclear energy are: rich resource in nature, less nuclear waste, low toxicity, nuclear non-proliferation and so on. Furthermore, China is a country with abundant thorium, thus it is necessary to develop the Thorium-based Molten Salt Reactor (TMSR) in China. Shanghai Institute of Applied Physics, Chinese Academy of Sciences (SINAP) had designed and constructed the first China's light-water reactor and developed a zero-power thorium-based molten salt reactor successfully in the early 1970s. The applied research project 'thorium molten salt reactor nuclear power system' by SINAP together with several other institutes had been accepted and granted by China government in 2011. The whole project has been divided into three stages: Firstly, built a 2 MW-zero-power high temperature solid molten salt reactor in 2015 and a 2 MW-zero-power high temperature liquid molten salt reactor in 2017. Secondly, in 2020 built a 10 MW high temperature liquid molten salt reactor. Thirdly, on the base of previous work, a 100 MW high temperature molten salt reactor should be achieving in 2030. After more than one years of efforts, a high quality scientific research team has been formed, which is able to design the molten salt reactor, the molten salt loop and related key equipment, the systems of molten salt preparation, purification and the radioactive gas removal. In the past one year, the initial physical design of high temperature molten salt reactor has been completed; the nuclear chemistry and radiation chemical laboratory has been built, a high temperature salt (HTS) loop and radioactive gas removal experiment device system have been successfully developed and constructed. Further, the preliminary study on reactor used carbon-carbon composite material has been investigated. (author)

  13. Opacity calculations for extreme physical systems: code RACHEL

    Science.gov (United States)

    Drska, Ladislav; Sinor, Milan

    1996-08-01

    Computer simulations of physical systems under extreme conditions (high density, temperature, etc.) require the availability of extensive sets of atomic data. This paper presents basic information on a self-consistent approach to calculations of radiative opacity, one of the key characteristics of such systems. After a short explanation of general concepts of the atomic physics of extreme systems, the structure of the opacity code RACHEL is discussed and some of its applications are presented.

  14. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Seo, K. W.

    2006-01-01

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC

  15. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W

    2006-01-15

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

  16. Treatment of plutonium process residues by molten salt oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J. [Los Alamos National Lab., NM (United States); Heslop, M. [Naval Surface Warfare Center (United States). Indian Head Div.; Wernly, K. [Molten Salt Oxidation Corp. (United States)

    1999-04-01

    Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible {sup 238}Pu material. Plutonium processing residues are injected into a molten salt bed with an excess of air. The salt (sodium carbonate) functions as a catalyst for the conversion of the organic material to carbon dioxide and water. Reactive species such as fluorine, chlorine, bromine, iodine, sulfur, phosphorous and arsenic in the organic waste react with the molten salt to form the corresponding neutralized salts, NaF, NaCl, NaBr, NaI, Na{sub 2}SO{sub 4}, Na{sub 3}PO{sub 4} and NaAsO{sub 2} or Na{sub 3}AsO4. Plutonium and other metals react with the molten salt and air to form metal salts or oxides. Saturated salt will be recycled and aqueous chemical separation will be used to recover the {sup 238}Pu. The Los Alamos National Laboratory system, which is currently in the conceptual design stage, will be scaled down from current systems for use inside a glovebox.

  17. Treatment of plutonium process residues by molten salt oxidation

    International Nuclear Information System (INIS)

    Stimmel, J.; Wishau, R.; Ramsey, K.B.; Montoya, A.; Brock, J.; Heslop, M.

    1999-01-01

    Molten Salt Oxidation (MSO) is a thermal process that can remove more than 99.999% of the organic matrix from combustible 238 Pu material. Plutonium processing residues are injected into a molten salt bed with an excess of air. The salt (sodium carbonate) functions as a catalyst for the conversion of the organic material to carbon dioxide and water. Reactive species such as fluorine, chlorine, bromine, iodine, sulfur, phosphorous and arsenic in the organic waste react with the molten salt to form the corresponding neutralized salts, NaF, NaCl, NaBr, NaI, Na 2 SO 4 , Na 3 PO 4 and NaAsO 2 or Na 3 AsO4. Plutonium and other metals react with the molten salt and air to form metal salts or oxides. Saturated salt will be recycled and aqueous chemical separation will be used to recover the 238 Pu. The Los Alamos National Laboratory system, which is currently in the conceptual design stage, will be scaled down from current systems for use inside a glovebox

  18. Coding and decoding for code division multiple user communication systems

    Science.gov (United States)

    Healy, T. J.

    1985-01-01

    A new algorithm is introduced which decodes code division multiple user communication signals. The algorithm makes use of the distinctive form or pattern of each signal to separate it from the composite signal created by the multiple users. Although the algorithm is presented in terms of frequency-hopped signals, the actual transmitter modulator can use any of the existing digital modulation techniques. The algorithm is applicable to error-free codes or to codes where controlled interference is permitted. It can be used when block synchronization is assumed, and in some cases when it is not. The paper also discusses briefly some of the codes which can be used in connection with the algorithm, and relates the algorithm to past studies which use other approaches to the same problem.

  19. Separate Turbo Code and Single Turbo Code Adaptive OFDM Transmissions

    Directory of Open Access Journals (Sweden)

    Lei Ye

    2009-01-01

    Full Text Available This paper discusses the application of adaptive modulation and adaptive rate turbo coding to orthogonal frequency-division multiplexing (OFDM, to increase throughput on the time and frequency selective channel. The adaptive turbo code scheme is based on a subband adaptive method, and compares two adaptive systems: a conventional approach where a separate turbo code is used for each subband, and a single turbo code adaptive system which uses a single turbo code over all subbands. Five modulation schemes (BPSK, QPSK, 8AMPM, 16QAM, and 64QAM are employed and turbo code rates considered are 1/2 and 1/3. The performances of both systems with high (10−2 and low (10−4 BER targets are compared. Simulation results for throughput and BER show that the single turbo code adaptive system provides a significant improvement.

  20. TASS/SMR Code Topical Report for SMART Plant, Vol. I: Code Structure, System Models, and Solution Methods

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Young Jong; Kim, Soo Hyoung; Kim, See Darl (and others)

    2008-10-15

    The TASS/SMR code has been developed with domestic technologies for the safety analysis of the SMART plant which is an integral type pressurized water reactor. It can be applied to the analysis of design basis accidents including non-LOCA (loss of coolant accident) and LOCA of the SMART plant. The TASS/SMR code can be applied to any plant regardless of the structural characteristics of a reactor since the code solves the same governing equations for both the primary and secondary system. The code has been developed to meet the requirements of the safety analysis code. This report describes the overall structure of the TASS/SMR, input processing, and the processes of a steady state and transient calculations. In addition, basic differential equations, finite difference equations, state relationships, and constitutive models are described in the report. First, the conservation equations, a discretization process for numerical analysis, search method for state relationship are described. Then, a core power model, heat transfer models, physical models for various components, and control and trip models are explained.